ML20027B586
| ML20027B586 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/21/1982 |
| From: | Vosbury F FRANKLIN INSTITUTE |
| To: | Peter Hearn NRC |
| Shared Package | |
| ML20027B587 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130 TAC-46855, TER-C5506-136, NUDOCS 8209230210 | |
| Download: ML20027B586 (24) | |
Text
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TECHNICAL EVALUATION REPORT PWR MAIN STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION (B-69)
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION NRC DOCKET NO. 50-312 FRC PROJECT C5506 NRCTAC NO. 46855 FRC ASSIGNM NTS NRC CONTRACT NO. NRC-03-81-130 FRC TASK 136 l
l Preparedby Franklin Research Center Author: F. Vosbury 20th and Race Street Philadelphia, PA 19103 FRC Group Leader:
R. C. Herrick Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: P. Hearn September 21, 1982 This report was prepared as an account of work sponsored t,y an agency of the United States Government. Neither theUnited States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or impiled, or assumes any legal liability,or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed In this report, or represents that its use by such third party would not infringe privately owned rights.
Prepared tiy:
Reviewed by:
Approved by:
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Principal Author: 6 Group Leader Dep'artment Director (Acting)
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.00. Franklin Research Center A Division cf The Franklin Institute The Ben >armn Franklin Parkway. PNia.. Pa. 19103(21S)448 100o kN DG11@ )
1 TER-C550 6-13 6 CONTENTS Section Title Page 1
INTRODUCTION.
1 1.1 Purpose of Review 1
1.2 Generic Background.
1 1.3 Plant-Specific Background 3
2 ACCEPTANCE CRITERIA.
4 3
TECHNICAL EVALUATION.
8 3.1 Review of Containment Pressure Response Analysis 8
3.2 Review of Reactivity Increase Analysis.
15 3.3 Review of Corrective Actions 18 4
CONCLUSIONS.
19 20 5
REFERENCES iii d A'J Franklin Research Center A w or n. r e us.uw.
TER-C550 6-13 6 FORDIORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operatina Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
Mr. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.
nklin Research Center A Duman of The Frerebn instante
9 TER-C550 6-13 6 1.
INTRODUCTION 1.1 PUKPOSE OF REVIEW This Technical Evaluation Report (TER) documents the a review of Sacramento Municipal Utility District's (SMUD) response to the Nuclear Regulatory Commission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break with Continued Feedwater Addition" (1), as it pertains to Rancho Seco Nuclear Generating Station Unit 1.
This evaluation was performed with the following objectived:
o to assess the conformance of SMUD's main steam line break (MSLB) analyses with the requirements of IE Bulletin 80-04 o to access SMUD's proposed interim and long-range corrective action plans and schedules, if needed as a result of the MSLB analyses.
1.2 GENERIC BACKGROUND In the summer of 1979, a pressurized water reactor (PWR) licensee submitted a report to the NRC that identified a deficiency in the plant's
(
original analysis of the containmeat pressurization resulting from a MSLB. A reanalysis of the containment pressure response following a MSL3 was performed, and it was determined daat, if the auxiliary feedwater (En() system continued to supply feedwater at runout conditions to the steam generator that had experienced the steam line break, containment design pressure would be exceeded in approximately 10 minutes. The long-term blowdown of the water supplied by the AFW system had not been considered in the earlier analysis.
On October 1,1979, the foregoing information was provided to all holders of operating licenses and construction permits as IE Information Notice 79-24 (2].
Another facility performed an accident analysis review pursuant to receipt of the information in the notice and discovered Ubat, with offsite electrical power available, the condensate pumps would feed the affected steam generator at an excessive rate. This excessive feed was not previously considered in the plant's analysis of a MSLB accident.
. b Franklin Research Center A Deveman of The Frannan ineenma
TER-C550 6-13 6 A third licensee informed the NRC of an error in the MSLB analysis for their plant. During a review of the MSLB analysis, for zero or low power at the end of core life, the licensee identified an incorrect postulation that the startup feedwater control valves would remain positioned "as is" during the transient. In reality, the startup feedwater control valves will ramp to 80% full open due to an override signal resulting from the low steam generator pressure reactor trip signal. Reanalysis of the events showed that opening of the startup valve and associated high feedwater addition to the affected staam generator would cause a rapid reactor cooldown and resultant reactor return-to-power response, a condition which is outside the plant design basis.
Because of these deficiencies identified in original MSLB accident analyses, the NRC issued IE Bulletin 80-04 on February 8,1980. This bulletin required all FIRS with operating licenses and certain near-term PWR operating license applicants to perform the following:
"1.
Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
2.
Review your analysis of the reactivity
<ese which results from a main steam line break inside or outside.untainment. This review should consider the reactor cooldown rate and Obe potential for the reactor to return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should includes a.
The boundary conditions for the analysis, e.g.,
tr.e end of life shutdown margin, the moderacor temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,
b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system,
. d!fEranklin Research Center A Dnamen of The Frannhainsstute
TER-C550 6-13 6 c.
The effect of extended water supply to the affected steam generator on the core criticality and return to power, d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum -
Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient.
3.
If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."
i 1.3 PLAIC-SPECIFIC BACKGROUND Sacramento Municipal Utility District responded to IE Bulletin 80-04 in a letter to the NRC dated May 6,1980 (3] and provided additional information in a letter dated May 28, 1982 (4]. The information in References 3 and 4 has been evaluated along with pertinent information from th: Rancho Seco Nuclear Generating Station Final Safety Analysis Report (FSAR) (5] to determine the adequacy of the Idcensee's compliance with IE Bulletin 80-04.
i i
I e Nhranklin Research Center A Dmmon of The Franssn insende
1 T ER-C550 6-13 6 2.
ACCEPTANCO CRITERIA The following criteria against which the Licensee's MSLB response was evaluated were provided by the NRC [6):
=
1.
PWR licensees' responses to IE Bulletin 80-04 shall include the following information related to their analysis of containment pressure and core reactivity response to a MSLB within or outside containment:
a.
A discussion of the continuation of flow to the affected steam generator, including dae, impact of runout flow from the AFW system and the impact of other energy sources, such as continuation of feedwater or condensate flow. AFW system runout flow should be determined from the manufacturer's pump curves at no backpressure, unless the system containe reliable anti-runout provisions or a more representative backpressure has been conservatively calculated. If l
a licensee assumes credit for anti-runout provisions, then i
justification and/or documentation used to determine that the provisions are reliable should be provided. Examples or devices for which provisions are reliable are anti-runout devices that use active components (e.g., automatically throttled valves) which meet the requirements of IEEE Std 279-1971 (7] and passive devices (e.g.,
flow orifices or savitating venturis),
b.
A determination of potential containment overpressure as a result of the impact of runout flow from the AFW system or the impact of other energy sources such as continuation of feedwater or condensate.
flow. Where a revised analysis is submitted or where reference is made to the existing FSAR analysis, the analysis,must show that runout AFW flow was included and that design containment pressure was not exceeded.
ge e rom con i ue f edwa add o du ing the M l
accident. Operator action to isolate AFW flow to the affected steam generator within the first 30 minutes of the start of the MSLB should be justified. If operator action is to be completed within the first 10 minutes, then the justification should address the indication available to the operator and the actions required.
Where operator action is required to prevent exceeding a design
- value, i.e., containment design pressure or specified acceptable fuel design limits, then the discussion should include the calculated time when the design value would be exceeded if no operator action were assumed. Where operator actions are performed between 10 and 30 minutes af ter the start of the MSLB, the justi-fication should address the indications available to the operator J
and the operator actions required, noting that for the first 30 minutes, all actions should be performed from the control room.
A nklin Research Center A Diesson of The Fransen inseewe i
TER-C550 6-13 6 d.
Where all water sources were not considered in the previous analysis, an indication should be provided of the core reactivity change which results from the inclusion of additional water sources. A submittal which does not determine the magnitude of reactivity change from an original analysis is not responsive to the requirements of IE Bulletin 80-04.
2.
If containment overpressure or a worsening of the reactor return-to-power with a violation of the specified acceptable fuel design limits described in Section 4.2 of the Standard Review Plan [8]
(i.e.,
increase in core reactivity) can occur by the licensee's analysis, the.lienesee shall provide the following additional information:
hheproposedcorrectiveactionstopreventcontainment a.
overpressure or the violation of fuel design limits, and the schedule for their completion b.
The interim actions that will be taken until the proposed corrective action is completed, if the unit is operating.
~
3.
The acceptable input assumptions used in the licensee's analysis of the core reactivity changes during a MSLB are given in Section 15.1.5 of the Standard Review Plan (9). The following specific assumptions should be used unless the analysis shows that a different assumption is more limiting:
Assumption II.3.b.:
Analysis should be performed to determine the most conservative assumption with respect to a loss of electrical power. A reactivity analysis should be conducted for a normal power situation as well as a loss of offsite power scenario, unless the licensee has previously conducted a sensitivity analysis which demonstrates that a particular assumption is more conservative.
Assumption II.3.d.:
The mcat restrictive single active failure in the safety injection system which has the effect of delaying the delivery of high concentration boric acid solution to the reactor coolant system, or any other single active failure affecting the plant response, should be considered.
Assumption II.3.g.:
The initial core flow should be chosen such that the post-MSLB shutdown margin is minimized (i.e., maximum initial core flow).
. Nb Franklin Research C.e.nter a om ono m. r wim.
i
TER-C550 6-13 6 The acceptable computer codes for t' e licensee's analysis of core reactivity changes are, by nuclear steam supply system (NSSS) vendor, the following CESEC (Combustion Engineering), IDFTRAN (Westinghouse),
and TRAP (Babcock & Wilcox). Other computer codes may be used, provided that these codes have previously been reviewed and found to be acceptable by the NRC staff. If a computer code is used which has not been revi wed, the licensee must describe the method employed to verify the code results in sufficient detail to permit the code to be reviewed for acceptability.
4.
If the AFW pumps can be damaged by extended operation at runout flow, the licensee's action to preclude damage should be reviewed for technical merit. Any active features should satisfy the requirements of IEEE Std 279-1971. Where no correctis, action has been proposed, 'this should be indicated to the NRC for further action and resolution.
5.
Modifications to the electrical instrumentation and controls needed to detect and initiate isolation of the affected steam generator and feedwater sources in order to prevsnt cantainment overpressure and/or i
unacceptable core reactivity increases must satisfy safety-grade requirements. Instrumentation that the operator relies upon to follow the accident and to determine isolation of the affected steam generator and feedwater sources should conform to the criteria contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors" (10), and the regulatory positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" (11].
l 6.
AFW system status should be reviewed to ensure that system heat removal capacity does not decrease below the minimum required level as a result i
of isolation of Obe affected steam generator and also that recent changes I
have not been made in the system which adversely affect vital assumptions of the containment pressure and core reactivity response analyses.
7.
The safety-grade requirements (redundancy, seismic and environmental qualifications, etc.) of the equipment that isolates the main feedwater (MFW) and AFW systems from the affected steam generator should be specified. The modifications of equipment that are relied upon to isolate the MFW and AFW systems from the affected steam generator should satisfy the following criteria to be considered safety-grade:
o Redundancy and power source requirements: The isolation valves should be designed to accommodate a single failure. A failure-modes-and-effects analysis should demonstrate that the system is capable of withstanding a single failure without loss of function.
The single failure analysis should be conducted in accordance with the appropriate rules of application of ANS-51.7/N658-1976, " Single Failure Criteria for PNR Fluid Systems" (12).
nklin Research Center A Dens.cn of The Frannan insatute
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TER-C550 6-13 6 o
Seismic requiren.ents: The isolation valves should be designed to Category I as recommended in Regulatory Guide 1.26 [13].
o Environmental qualification: The isolation valves should satisfy the requirements of NURE.G-0588, Pav.1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" (14 ].
o Quality standards: The isolation valves should satisfy Group B qaality atandards as recommended in Regulatory Guide 1.26 or similar quality standards from the plant's licensing bases.
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. A Ub nklin Research Center
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TER-C550 6-13 6 3.
TECHNICAL EVALUATION Under contract to the' NBC, the scope of work included the following:
1.
Review the Licensee's response to IE Bulletin 80-04 against the acceptance criteria.
2.
a.
Evaluate the Licensee's MSLB analyses for the potential of overpressurizing the containment and with respect to the core reactivity increase due to the effect of continued feedwater flow.
b.
Evaluate the Licensee's proposed corrective actions and schedule for implementation if the findings of Task 2a indicate that a potential exists for overpressurizing the containment or worsening the reactor return-to-power in the event of _a MSLB accident.
3.
Prepare a TER for each plant based on the evaluation of the information presented for Tasks 1 and 2 above.
This report constitutes a TER in satisfaction of Task 3.
Sections 3.1 through 3.3 of this report state the requirements of IE Bulletin 80-04 by subsection, summarize the Licensee's statements and conclusions regarding these requirements, and present a discussion of the Licensee's evaluation followed by conclusions and recommendations.
3.1 REVIEW OF CONTAINMENT PRESSURE RESPONSE ANALYSIS The requirement from IE Bulletin 80-04, Item 1, is as follows:
(
" Review the containment pressure response analysis to determine if the l
potential for containment overpressure for a main steam line break inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow. In your review, consider l
your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow."
l 3.1.1 Summary of Licensee Statements and Cenclusions In regard to the review of the containment pressure response analysis, the i
Licensee stated [3]:
i e
nklin Research Center A Dnnsoon of The Franen Insetute l
l - - - -
o TER-C550 6-13 6 "To assess the impact of the steam line break and the ability of plant systems and structures to respond in a safe manner to this type of transient, the following review was conducted.
4 A review of the Rancho Seco Unit 1 Accident Analysis, FSAR Chapter 14.2, shows that the plant systems and structures are designed to respond to a double ended guillotine steam line break inside of the containment.
As a result of the reactor trip, the turbine stop valves close, isolating the steam side of the unaffected steam generator. The reactor trip also results in the isolatior, of main feedwater to both steam generators.
i Additionally, a feedwater isolation signal to the affected steam generator is provided by low steam line pressure.
The auxiliary feedwater pumps will automatically start on the loss of both main feedwater pumps, providing feedwater to both steam generators to maintain a two foot minimum downcomer water level in the steam generators. The affected steam generator would be identified and auxiliary feedwater isolated to the ganerator in accordance with established and approved Emergency Procedures (D.13).
Once the affected steam generator is isolated, it will blow dry. The reactor coolant system would be cooled by venting steam from the unaffected steam generator to the condenser or atmosphere. The operator would then systematically cool and depressurize by controlled auxiliary feedwater addition and steam venting.
As a result of the mass and energy released to the containment, the internal pressure would increase to approximately 30 psig."
In response to a question regarding operator action af ter a MSLB, the Licensee stated [4} :
"Following a main steam line break the control room operator would identify the affected steam generator by observing and comparing steam generator level and/or pressure indications in the control room.
Auxiliary feedwater flow to the affected steam generator would be isolated by the closure of one valve which is operable from the control room. These steps are listed as "Immediate Operator Actions" in the emergency operating procedure, " Steam Line/ Feed Line Rupture". These operations are straight forward, well understood by the operators; all occur in the control room; and would certainly occur within a ten minute
~
period."
In regard to the ability of the AFW pumps to remain operable during a MSLB, the Licensee stated (4):
"The District is unable to provide information concerning the ability of the auxiliary feedwater pumps to continue to run without damage at runout flow for thirty minutes. Extensive test data would have to be developed j
4 ddu Franklin Research Center A Dmeson of The Franadwi esttute
o TER-C550 6-13 6 for such an evaluation, and (since an upgraded auxiliary feedwater initiation and control system is being implemented at Rancho Seco to provide the Class 1 control of AFW flow under the conditions of MSLB] we feel this level of effort is not justified."
3.1.2 Evaluation The Licensee's submittais [3, 4] concerning the containment pressure response following a MSLB a.1d applicable sections of the Rancho Seco Nuclear Generating Station FSAR ['2] were reviewed in order to evaluate whether dhe following portions of rie acceptance criteria were met:
o Criterion 1.a - Continuation of flow to the affected steam generator o Criterion 1.b - Potential for containment overpressure o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 4 -
Potential for AFW pump damage o Criterion 5 -
Design of steam and feedwater isolation system o Criterion 6 -
Decay heat removal capacity o Criterion 7 -
Safety-grade requirements for MFW and AFW isolation valves.
Rancho Seco Nuclear Generating Station is a Babcock and Wilcox-designed, two-loop, 2772-MWt plant.
In the event of a MSLB, the following systems actuate to provide necessary protection:
o Reactor trip on high flux o The reactor trip signals closure of:
a.
turbine stop valves b.
MFW control valves and startup control valves to each steam generator c.
MFW stop valves (main feedwater is totally isolated 26 seconds af ter the reactor trip)
N_nklin Research._ Center
=
TER-C550 6-13 6 o low steam line pressure also signals the closure of MFW control valves and startup control valves o AFW pumps start on loss of both MFW pumps (steam driven) and provide feedwater to both steam generators to maintain a two foot downcomer water level.
o AFW flow is isolated to the affected steam generator within 10 minutes.
The FSAR analysis (Section 14.2.2.1.3.2) also nede the following assumptions "For a steam line break with a concurrent failure of one turbine stop valve in the unbroken line, the back pressure in the unbroken steam line can prevent dhe turbine stop valves in the broken line from seating. The remaining turbine stop valve in the unbroken line will not open due to the forward pressure exerted on it.
However, for this analysis all the turbine stop valves are assumed to open resulting in a continuous double blowdown through the broken line." (See Figure 3-1)
The double blowdown of the two steam generators produces a peak containment pressure of 56 psig occurring at about 65 seconds. Containment design pressure is 59 psig. The blowdown of a single steam generator under the same initial conditions produced a peak pressure of 31 psig occurring at 65 seconds.
The operator has sufficient instrumentation to analyze the type of accident that has occurred. Once the operator has determined what type of accident has occurred, he has minimal actions to perform (closure of one valve operable from the control room) to isolate feedwater to the affected steam generator.
On November 17, 1980 (15), SMUD provided details of a safety-grade, automatically initiated, AFW system design to feed only the unaffected steam generatar in tt-event of a MSLB. The Licensee committed to install this system in the latter part of 1982. The final design of this AFW initiation system was provided in Reference 16.
l A review of References 15 and 16 determined that the proposed emergency feed initiation and control system (EFIC) is an instrumentation system designed to provide the following:
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TER-C550 6-13 6 e initiation of auxiliary feedwater (APW)'
o control of APW at appropriate setpoints (approxmately 3, 20, and 31.5 feet) o level rate control when required to minimize overcooling o isolation ~of the main steam and main feedwater lines of a depressurized steam generator o the selection of the appropriate steam generator (s) under conditions of steamline break or main feedwater or emergency feedwater line break downstream of the last check valve o termination of main feedwater to a steam generator on approach to overfill conditions o termination of AFW to a steam generator on approach to overfill conditions o control of atmospheric dump valves to a predetermined setpoint.
The EFIC logic issues a call for AFW auto-initiation when o all four reactor coolant pumps are tripped o both main feedwater pumps are tripped (i.e., low discharge pressure) o the level of either steam generator is low 4
o either steam generator pressure is low o flux to MFW' flow ratio trip is present.
Other functions of the EFIC logic ares o Issues a call for steam generator (SG) A main feedwater and main steamline isolation when SG A pressure is low o Issues a call for SG B main feedwater and main steamline isolation when SG B pressure is low o Signals approach to SG A overfill when SG A level exceeds a high level setpoint o signals approach to SG B overfill when SG B level exceeds a high level setpoint o Provides for manually initiated individual shutdown bypassing of reactor coolant pumps, main feedwater pumps, and SG pressure initiation.
b Franklin Research Center A Onnson of The Frannan Instema e
TER,-C550 6-13 6 of AFW as a function of permissive conditions. The bypass (es) are automatically removed when the permissive condition terminates o Provides for maintenance bypassing of an EFIC initiate logic.
In the event of a steam line break or feed line break, the EFIC system is designed to isolate the steam and feedwater lines and to provide auxiliary feedwater to the intact steam generator. The system is designed such that no single active failure will either prevent auxiliary feedwater from being supplied to the intact steam generator or allow auxiliary feedwater to be supplied to the broken steam generator.
To meet the requirements for steam line or feed line break protection, the following design was implemented:
o Isolation - low steam pressure (below approximately 600 psig) in either SG will isolate the main steamlines and main feedwater line to the affected SG
~
o SG selection -
a.
If both SGs are above 600 psig, supply AW to both SGs b.
If one SG is below 600 psig, supply AFW to the other SG c.
If both SGs are below 600 psig but the pressure difference between the two SGs exceeds a fixed setpoint (approximately 150 psig) supply AFW only to the SG with the higher pressure d.
If both SGs are below 600 psig and the pressure difference is less than die fixed setpoint, supply AFW to both SGs The EFIC system was designed to safety-grade and IEEE Std 279-1971 requirements.
The environmental qualification of safety-related electrical and l
mechanical components is being reviewed separately by the NRC and is not within the scope of this review.
The review did not determine if the instrumentation that the operator relies upon to follow the accident and isolate the affected steam generator conforms to the criteria in ANS/ ANSI-4.5-1980 (10] and Regulatory Guide 1.97 [11].
M Franklin Research Center A Dresson of The Franda insoture
a TER-C550 6-13 6 In the event of a MSLB with a concurrent single active failure of the turbua.tcp valve, both steam generators will blow down. The two generators will depressurize, and since the differential pressure between the two steam generatcrs will be less than 150 psig, the EFIC system will cause both steam generators to be supplied with AFW (case d above), possibly producing a more severe pressure transient than was previously analyzed.
In the event of a double blowdown, no runout protection would be in effect; both pumps would operate at runout flow and could possibly be subject to damage.
3.1.3 Conclusion and Recommendations The Licensee's responses [3, 4] and the Rancho Seco Nuclear Generating Station FSAR [5] adequately address the concerns of Item 1 of IE Bulletin 80-04.
The proposed EFIC system will provide safety-grade protection against a MSLB and eliminate the need for operator action except in the case of the double blowdown which may produce a more severe transient than that previously analyzed in the FSAR.
The Licensee should provide a full spectrum analysis of the containment pressure response to a MSLB prior to placing the EFIC system in operation.
The AFW pumps are not protected against runout flow in the current AFW system and may experience damage. The safety-grade EFIC system will prevent the AFW pumps from experiencing runout conditions, except during a double steam generator blowdown with the EFIC system installed.
3.2 REVIEW OF REACTIVITY INCREASE ANALYSIS The requirement from IE Bulletin 80-04, Item 2, is as follows:
" Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider dae reactor cooldown rate and the potential for the reactor to return-to-power with the most reactive control rod in the fully withdrawn position.
If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated, the report of this review should include:
. OUb&
Franklin Research Center A Onmon of The Fratmhnineen,te
o TER-C550 6-13 6 a.
The boundary conditions for the analysis, e.g.,
the end of life shutdown margin, the moderator temperature coefficient, power level
^
and the not effect of the associated steam generator water inventory on the reactor system cooling, etc.,
b.
The most restrictive single active failure in the safety injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, c.
The effect of extended water supply to the affected steam generator on the core criticality and return-to-power, d.
The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyze 1 transient."
+
3.2.1 Summary of Licansee Statements and Conclusions
~
In regard to the reactivity increase resulting from a MSLB with continued feedwater addition, the Licensee stated (3):
"A review of the Rancho.Seco Unit l' Accident Analysis, FSAR Chapter 14.2, shows that the plant systems and structures are designed to respond to a double-ended guillotine steam line break inside of the containment. As discussed in Section 14.2.2.1.3.1, Accident Dynamics, a main steam line break would lead to a rapid decrease in secondary steam pressure and an increase in steam flow. This increased steam flow increases the primary to secondary heat transfer, thus lowering the primary coolant pressure i
and average temperature. Due to the large negative moderator temperature coefficient, the reactor thermal power will increase resulting in a reactor trip at 106 percent on high neutron flux in about 10 seconds.
As a result of the reactor trip, the turbine stop valves close, isolating the steam side of the unaffected steam generator. The reactor trip also-i results in the isolation of main feedwater to both steam generators.
Additionally, a feedwater isolation signal to the affected steam generator is provided by low steam line pressure.
The auxiliary feedwater pumps will automatically start on the loss of both main feedwater pumps, providing feedwater to botu steam generators to maintain a two foot minimum downcomer water level in the steam generators. The affected steam generator would be identified and auxiliary feedwater isolated to the generr. tor in accordance with established and approved Emergency Procedures (D.13).
Once the affected steam generator is isolated, it will blow dry.
At 4 psig high containment pressure or 1600 psig low primary system pressure, the high, pressure injection system will actuate if not actuated,
s b Franklin Research C. enter 4om.a.e n.Frn.enwi.nu
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TER-C550 6-13 6 manually by the operator. The water added by the high pressure injection system is from the borated water storage tank (BWST), which contains an equivalent of 390,000 gallons of 1800 ppm borated water. By the addition of this borated water, adequate shutdown margin would be provided to prevent recriticality."
3.2.2 Evaluation The Licensee's analysis of the core reactivity increase resulting from a MSLB with continued feedwater addition was reviewed in order to evaluate whether the following acceptance criteria were mets o Criterion 1.c - Ability to detect and isolate the damaged steam generator o Criterion 1.d - Changes in core reactivity increase o Criterion 3 Analysis assumptions.
The FSAR analysis of the reactivity increase resulting from a MSLB and Reference 3 were reviewed. From that review, it was determined that the analysis is conservative in its assumptio,ns and that the assumptions are in accordance with those in Acceptance Criterion 3.
In the worst case MSLB, which assumes full power conditions, a double-ended rupture at the steam generator exit, and concurrent failure of a turbine stop valve in the unbroken line (produci'ng a double blowdown), a peak power of 106% occurs at 6 seconds, at which time a high flux reactor trip occurs, inserting the control rods. Af ter the reactor trip the core remains subcritical. The minimum suberitical margin attained af ter the trip is 0.7%.
The predicted core transient did not result in a violation of the specified acceptable fuel design limits.
However, for the proposed EFIC system, the concurrent failure of a turbine stop valve in the unbroken line causes both steam generators to blow down. The EFIC system directs AFW flow to both steam generators, causing additional cooldown and potentially leading to a lessening of the subcritical margin and a return-to-power.
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3.2.3 Conclusion and Recoinmendation For the current plant design, the Licensee's responses (3, 4] and FSAR adequately address the concerns of Item 2 of IE Bulletin 80-04.
All potential sources of water were identified, no return-to-power is predicted, and there is no violation of the specified acceptable fuel design limits; the FSAR analysis of the reactivity increase resulting from a MSLB remains vnlid.
However, prior to-the installation of the EFIC system, the Licensee should perform an analysis of the reactivity response to a MSLB to ensure that an unacceptable worsening ofia return-to-power does not occur and that the specified acceptable fuel design limits are not exceeded.
3.3 REVIEW OF CORRECTIVE ACTIONS The requirement from IE Bulletin 80-04, Item 3, is as follows:
"If the potential for containment overpressure exists or the reactor return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that will.be taken until the proposed corrective action is completed."
3.3.1 Summarv of Licensee Statements and Conclusions The Licensee stated (3):
"The attached response provides a description and results of the review.
Based on these results, the District is assured that a main steam line break inside containment will not result in over pressurization of the Containment Building, and the cooldown rate of the primary system will be within the limits of the technical specifications, and the reactor will remain subcritical."
3.3.2 Evaluation, Conclusion, and Recommendations The Licensee is required to provide an analysis of a MSLB to determine the containment pressure response and reactivity increase for the proposed EFIC system since the feeding of both steam generators is not bounded by the assumptions in the FSAR analysis.
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CONCLUSIONS with respect to the Rancho Seco Nuclear Generating Station, conclusions regarding Sacramento Municipal Utility District's response to IE Bulletin 80-04 are as follows:
o There is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition under the current AN system design.
o An analysis of the containment pressure response to a MSLB should be conducted for the EFIC system.
o The FSAR reactivity increase analysis bounds the current plant design.
o An analysis of the reactivity response to a MSLB should be conducted for the EFIC system.
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o Under the current plant design, the AN pumps are not protected against runout flow, o Under the design of the EFIC system, the AN pumps are protected against runout except during a double steam generator blowdown.
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REFERENCES 1.
" Analysis of a PWR Main Steam L'ne Break with Continued Feedwater Additi' n" o
NBC Office of Inspection and Enforcement, February 8,1980 IE Bulletin 80-04 2.
"Overpressurization of the Containment of a PWR Plant af ter a Main Steam Line Break" NRC Office of Inspection and Enforcement, Octot,er 1,1979 IE Information Notice 79-24 3.
J. J. Mattimoe (SMUD)
Letter to R. H. Engelken (NRC, Region V)
Subject:
IE Bulletin 80-04 May 6,1980 4.
J. J. Mattimoe (SMUD)
Letter to J. F. Stolz (NRR)
Subject:
Response to Hequest for Additional Information, IE Bulletin 80-04 May 28,1982 5.
Rancho Seco Nuclear Generating Station Final Safety Analys.is Report, through Amendment 27 Sacramento Municipal Utility District 6.
Technical Evaluation Report "PWR Main Steam Line Break with Continued Feedwater Addition - Review of Acceptance Criteria" Franklin Research Center, November 17, 1981 TER-C550 6-119 7.
"Criter!a for Protection Systems for Nuclear' Power Generating Stations" Institute of Electrical and Electronics Engineers, New York, NY, 1971 IEEE Std 279-1971 8.
Standard Review Plan, Section 4.2
" Fuel System Design" NBC, July 1981 NUREG-0800 9.
Standard Review Plan, Section 15.1.5
" Steam System Piping Failures Inside and Outside of Containment (PWR) "
NRC, July 1981 NUREG-0800
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" Criteria for Accider.t Manitoring Functions in Light-Water-Cooled j
Reactors" American Nuclear Society, Hinsdale, IL, December 1980 ANS/ ANSI-4.5-1980 11.
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" Rev. 2 NBC, December 1980 Regulatory Guide 1.97 12.
" Single Failure Criteria for PWR Fluid Systems" American Nuclear Society, Hinsdale, IL, June 1976 ANS-51.7/N658-1976 13. " Quality Group Classifications and Standards for Water, Steam, ar.d Radioactive-Waste-Containing Components of Nuclear Power Plants" Rev. 3 NRC, February 1976 Regulatory Guide 1.26 14.
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" Rev. 1 NBC, July 1981 NUREG-0588 15.
W. S. Bossenmaier (SMUD)
Letter to R. W. Reid (NRR)
Subject Upgraded Auxiliary Feedwater System Design Submittal November 17, 1980 16.
J. J. Mattimoe (SMUD)
Letter to J. F. Stolz (NRR)
Subject:
'IMI Action Plan Items II.E.1.2 and II.K.2.10 September 8,1981
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