ML20153H556
| ML20153H556 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/04/1988 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 1520, NUDOCS 8805130059 | |
| Download: ML20153H556 (12) | |
Text
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TOLEDO
%mm EDISON A Certertr Crergy Cm DONAU) C. SHELTON Vre Presert-Nucirar Docket No. 50-346
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License No. NPF-3 Serial No. 1520 May 4, 1988 United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C.
20555 Subj ect :
Additional Information Regarding License Amendment Application to Revise Main Steam Safety Valve Relief Capacity /High Flux Trip Setpoint and Restate ASME Code Requirements for Hain Steam Safety Valves (TAC No. 67394)
Gentlemen:
During a March 17, 1988, meeting with Mr.. A. V. DeAgazio, NRR Proj ect Manager for Davis-Besse Nuclear Pover Station (DBNPS), Unit 1 and members of the NRC Staff, Toledo Edison (TED) provided a presentation and information regarding the subject License Amendment application as previously submitted on March 4, 1988 (Serial No. 1487).
During this meeting TED vas requested to provide additional information with respect to the impact of the proposed changes on the present Updated Safety Analysis Report (USAR) Chapter 15 Accident Analyses.
Attached please find information which further assesses the impact of the proposed change as requested.
The results of this assessment reconfirm the conclusions presented in Serial No. 1487.
Your expeditious review and approval consistent with the previously requested June 1, 1988 License Amendment issuance date is herewith reiterated.
If you have any further questions, please contact Mr. R. V. Schrauder, Nuclear Licensing Manager, at (419) 249-2366.
Very tr
. yours, HCitlt j
Q Attachment 0
cc: A. B. Davis, Regional Administrator DB-1 Resident Inspector THE TOLEDO FOSON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OH!O 43652 8805130059 880504 PDR ADOCK 05000346 (p
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Docket No. 50-346 License No. NPP-3 Serial No. 1520 Enclosure ADDITIONAL INFORMATION REGARDING APPLICATION FOR AMENDHENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POVER STATION UNIT NO. 1 Attached is additional information to support issuance of requested changes to the Davis-Besse Nuclear Power Station, Unit No. 1, Facility Opereting License No. NPF-3, as previously submitted by Serial No. 1487, dated March 4, 1988.
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By h
E'C'.
Shelton, Vice Pre'sideht, Nuclear Sworn and subscribed before me this 4th day of May, 1988.
Nam i e hrwrrOedw/
Notary E6blic, State of Ohio My commission expires NANCY L DA'/SCHacDip, pe;3y g g UllAWA C0iHHY, OHl0 My Cennissiori Errirss Oct 8, it;;
i Dockot No. 50-346 l
Liernsa No. NPF-3 Serial No. 1520 Attachment Page 1 Question:
The proposed Technical Specification change presented in Serial No. 1487 dated March 4, 1988 vill replace the present staggered HSSV setpoints with vording that will allow all but one MSSV per SG header to be set as high as 1100 psig.
Provide additional supporting information for this change by further assessing its impact on the present Chapter 15 Accident Analyses.
Answer:
The purpose of the MSSVs is to provide overpressure protection for the secondary side of the steam generators (SG) and the Main Steam (MS) System.
During normal plant operation the MSSVs are closed and the turbine-condenser combination provides the heat sink for the RCS. The MSSVs are only utilized following a turbine trip when the steam removal path provided by the turbine is lost. This causes the MS System to pressurize until the MSSV setpoints are reached and sufficient steam release through the MSSVs is obtained to maintain MS System pressure at or below the MSSV setpoint. Actuation of HSSVs establishes another path for SG heat removal by the venting of steam directly to the atmosphere.
The heat sink temperature for the RCS that is established by the SGs corresponds to the saturation temperature associated with the SG pressure. The increase in SG pressure following a turbine trip up to the 1050 - 1100 psig setpoint range of the MSSVs causes the heat sink temperature for the RCS to increase from - 530'F before the turbine trip to the 550 to 560*f range following the turbine trip.
If a reactor trip has not occurred in conjunction with the turbine trip, this increase in the RCS heat sink temperature could cause a transient increase in average coolant temperature and pressure.
If this increase in temperature and ptessure exceeds the normal operating envelope of the RCS, the Reactor Protection System (RPS) trips the reactor. This terminates the RCS temperature and pressure excursion.
If a reactor trip has occurred in conjunction with the turbine trip, the increase in heat sink temperature to the 550 to 560'F range establishes the average RCS temperature corresponding to this heat sink temperature.
Since the reactor core decay heat decreases with time following a reactor trip, within approximately one minute the decay heat load is lov enough that only the first bank of MSSVs vould be needed to match the RCS decay heat.
As noted in the TED submittal of March 4, 1988 (Serial No. 1487), this first bank of HSSVs is made up of a minimum of two valves per header set at 1050 psig. After this time, the maximum RCS heat sink temperature is determined by the setpoint of this first bank (lovest set) of HSSVs. Also, as noted in Serial No. 1487, the ASHE code requires that at least one safety valve be set at the system design pressure (1050 psig). This setpoint is unaffected by the proposed changes.
It is noted that to ensure adequate long term cooling of the RCS by use of the
Dockit No. 50-346 Liesnse No. NPF-3 Ssrial No. 1520 Attachment Page 2 SGs, it is only necessary that the MSSVs be able to vent sufficient steam to match the RCS decay heat load. The proposed changes ensure that overpressure protection of the SG and MS System is being provided. Variation of the exact setpoint pressure for this steam venting, in the range from 1050 psig to 1100 psig, has a minimal and acceptable effect on the RCS.
The Updated Safety Analysis Report (USAR) Chapter 15 Accident Analyses demonstrate that the plant is designed to provide the required degree of protection for the RCS and to limit off-site doses to within 10CFR100 limits.
Consequently, the accidents and abnormal transients investigated in Chapter 15 of the USAR must satisfy appropriate acceptance criteria.
Although the MSSVs are included in the analytical models used for certain accidents, their setpoints are not important for satisfying the appropriate acceptance criteria as evaluated below.
The acceptance criteria specified for protection of the RCS in Chapter 15 accidents are typically:
- Thermal power f 112% of rated power
- RCS pressure f code pressure limits
- Minimum DNB ratio of 2 1.3 The MSSVs are not required to satisfy any of the above criteria. The Reactor Protection System (RPS) ensures that the above criteria are satisfied by tripping the reactor on appropriate RCS conditions, typically on either high neutron flux or high RCS pressure.
Once the reactor is tripped, the reactor power rapidly decays, thereby ensuring the thermal power criteria is satisfied. The thermal power limitation of 112% ensures that a departure from nucleate boiling (DNB) heat transfer condition does not occur in the core.
This provides protection to the fuel cladding by limiting the fuel cladding temperature to approximately the core coolant temperature.
Since the peak thermal power typically occurs immediately following the reactor trip and before MSSVs are actuated, the MSSVs have no impact upon satisfying this acceptance criteria.
The reactor trip, in conjunction with the pressurizer code safety valves, is utilized to ensure that the RCS pressure remains within code limits.
The MSSV setpoint may affect the initiation of steaming from the secondary side, but it has no measurable impact upon the ability of the plant to satisfy the acceptance criteria for RCS pressure limits.
For any event, RCS pressure is maintained within code limits by the presence of the pressurizer code safety valves.
The ability of the plant to satisfy 10CFR100 dose limits is also unaffected by MSSV setpoint changes.
The setpoint change has some impact upon the timing of valve actuation and steaming of the secondary side in the first few seconds following a turbine trip, and also affects the secondary side
Dockst No. 50-346 Licsnsa No. NPF-3 Serial No. 1520 Attachment Page 3 pressure response during the first minute following a reactor trip.
As noted above, after approximately 1 minute the reactor decay heat is lov enough that only the first bank of HSSVs set at 1050 psig is required to match decay heat.
Although the timing of actuation and pressure response for the MS System can initially be affected by MSSV setpoints, the integrated amount of steam released to the atmosphere through the MSSVs is essentially unaffected.
The secondary side of the SG provides an integrated heat removal enpability that is dependent upon the mass inventory in the SGs.
Since the integrated energy released by the primary system over time and the initial SG inventory are basically independent of MSSV setpoint, the integrated steaming of water mass through the MSSV is also independent of the MSSV setpoints.
Consequently, off-site doses due to mass release through the MSSVs are unaffected by the MSSV setpoints.
A list of Chapter 15 accidents that include modeling of the steam generator secondary side where HSSV actuation can occur is presented in Table 1.
This Table also identifies the accident acceptance criteria and describes how HSSV setpoints impact the response of the plant to satisfy the specified acceptance criteria.
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Dock 2t No. 50-346 Lic2nsa No. NPF-3 S rial No. 1520 Attachment Page 4 Table 1 - List of Chapter 15 Accidents That Utilize HSG'!s Section Accident Acceptance Criteria Evaluation 15.2.2 Uncontrolled i) Thermal power Reactor trip on CRA Group-f 112% of rated either high RCS Vithdrawal at power.
pressure or high Power
- 11) RCS pressure does neutron flux not exceed code terminates pressure limits.
Since peak values occur before MSSVs vould actuate, HSSVs are not modeled for this transient.
15.2.5 Loss of Forced i) Minimum DNB ratio Reactor trip by RC Pump Flov i 1.30 for pump RPS terminates the coastdovn transient.
Immediately
- 11) No fuel cladding following reactor failure for trip DNB ratio locked rotor increases with accident.
decreasing neutron power.
Since minimum values of DNB ratio occur before MSSV actuation occurs, HSSVs are not modeled for this transient.
15.2.7 Loss of i) No fuel damage.
For transients External
- 11) RCS pressure does initiated at high Load / Turbine not exceed code power levels Trip pressure limits.
where the plant can no longer successfully runback, a reactor trip occurs on high RCS pressure.
This reactor trip causes reactor power to rapidly decrease providing protection against fuel damage.
RCS pressure protection is provided by the decrease in reactor power and the
4 Dockst N3. 50-346 Lic;nsa No. NPF-3 Serial No. 1520 Attachment Page 5 Section Accident Acceptance Criteria Evaluation pressurizer code safety valves. The MSSVs provide the steaming path for heat removal by the SGs. The change in MSSV setpoints does not prevent secondary side stenming from being established.
15.2.8 Loss of i) No fuel damage.
As with the Feedvater
reactor trip rn high RCS pressure in combination with pressurizer code safety valvet is utilized to satisfy the accident acceptance l
criteria. The secondary side provides the long term l
cooling for l
the RCS via l
steaming through the MSSVs.
The MSSV setpoint change does not prevent this heat sink from being established.
Vithin one minute decay heat is lov enough for the first bank of HSSVs at 1050 psig to accommodate the steam demand. The 1050 psig setpoint for the first bank of MSSVs is unaffected by this proposed change because it is an ASME code re-quirement.
I
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Dockot No. 50-346 Lic1nsa No. NPF-3 Sericl No. 1520 Attachment Page 6 Section Accident Acceptance Criteria Evaluation 15.2.9 Loss of AC i) No fuel damage.
Reactor trip occurs Power to
- 11) RCS pressure with loss of Station does not exceed station power.
Auxiliaries code presstre Auxiliary (Station limits.
Feedvater (AFV)
Blackout) flov establishes natural circulation cooling of RCS.
MSSVs provide the steaming path from SGs. Ability to steam the SG and establish AFV flow is unaffected by MSSV setpoints.
Vithin one minute SG pressure is at the 1050 psig pressure associated with first bank of MSSVs due to the reduced decay heat.
The 1050 psig setpoint for the first bank of HSSVs is unaffected by this proposed change because it is an ASME code requirement.
15.2.10 Excessive Heat i) No fuel damage.
RCS seceptance Removal due
- 11) RCS pressure does criteria are to Feedvater not exceed code satisfied by a Halfunction pressure limits.
reactor trip on high flux if required.
MSSVs are not used to mitigate primary side response.
Even vith the change in HSSV setpoints sufficient relieving capacity is available to prevent overpressurizing the secondary side.
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Docket N).-50-346 Lic nsa Ns. NPF-3 Serial No. 1520 Attachment Page 7 Section Accident Acceptance Criteria Evaluation 15.2.11 Excessive Load i) Core cooling Events of this Increase remains intact.
type are either
- 11) No SG tube failure controlled by the to cause loss Integrated Control of RCS pressure System (ICS) integrity.
without causing a lii) Doses are within reactor trip and 10CFR100 limits.
actuation of MSSVs or are bounded by steam line breaks.
See 15.4.4 evaluation for steam line breaks.
15.3.1 Small break i) No core damage.
Small break LOCAs LOCAs that utilize the SGs are long term transients
(> 1000 sec). The impact of revised MSSV settings on initial pressure response of SG during first 4
minute following reactor trip has no noticeable impact on the total heat i
removal capability of the SGs during this transient.
Af ter - 1 minute decay heat is lov enough for MSSVs at 1050 psig to provide necessary steaming.
15.3.2 Minor Secondary These accidents are System Pipe bounded by steam Ruptures line breaks discussed in Section 15.4.4.
15.4.2 Steam Gener-
slightly decrease
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boeketN2. 50-346 Liesns2 No. NPF-3 S; rial No. 1520 Attachment Page 8 Section Accident Acceptance Criteria Evaluation
- 11) No loss of RCS inventory release pressure boundary off-site by integrity due to initially secondary side decreasing 6P pressure and between the primary resultant and secondary temperature sides.
Impact gradients is negligible on causing a SG off-site release.
tube failure.
Analyses conservatively assumed a constant leak flov from the primary to secondary independent of pressure. Change in HSSV setpoint does not impose any limiting thermal stresses on SG tubes.
13.4.4 Steam Line
- 1) The core remains change in HSSV Break intact for setpoint has no effective cooling.
impact upon
- 11) N9 loss of RCS Chapter 15 Main pressure boundary Steam Line Break integrity due to (HSLB) Analysis, loss of secondary A HSL6 causes side pressure and secondary side resultant depressurization temperature and resultant gradients causing cooldown of RCS.
a SG tube failure.
MSSVs are only lii) Doses are within actuated following 10CFR100 limits, isolation of the unaffected SG vhich terminates its depressurization.
Subsequent re-pressurization of the unaffected SG is dependent upon RCS conditions resulting from continued blovdown of the affected generator. Any actuation of
Dock 9t No. 50-346 Lic nsa No. NPF-3 m
k-S; riel No. 1520 Attachment Page 9 Section Accident Acceptance Criteria Evaluation HSSVs that vould
{
occur following a HSLB vould only lift the first bank of HSSVs at 1050 psig.
I Change in HSSV setpoint has no impact on transient results shown in Chapter 15.
15.4.5 Break in
- 1) Doses are within Limiting accident Instrument 10CFR100 limits.
in Chapter 15 is a Line or line letdown line from Primary rupture outside System that containment. This Penetrates break is assumed to containment be terminated by Safety Features Actuation System (SFAS) closure of the letdown line isolation valve upon lov RCS l
pressure of 1600 l
psig.
Since the accident involves a reactor trip and is terminated after reactor trip by a SFAS signal, steam generator heat remaval following the reactor trip is required.
Change in MSSV setpoint i
vould slightly delay initiation of HSSV steaming, and consequently, could cause a slight increase in RCS temperature following reactor trip.
Although this effect could slightly delay
'Dockst No 50 346 Licinso No. NPF-3 Serial No. 1520 Attachment Page:10 Section Accident Acceptance Criteria Evaluation isolation of the break, there is no significant effect upon off-site doses and the 10CFR100 dose criteria for this accident would still be satisfied.
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