ML20011E809
| ML20011E809 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 02/07/1990 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9002220518 | |
| Download: ML20011E809 (9) | |
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,14' BALTI M O RC-3
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~ CHARLES CENTER P.O. BOX 1476 BALTIMORE, MARYLAND 21203 q
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(so0 soo amss U. S.' Nuclear' Regulatory Commission Washington, DC. 20555 ATTENTION:
Document Control Desk H
SUBJECT:
- Calvert: Cliffs Nuclear Power Plant Unit No.1; Docket Nos. 50-317 Description of Calvert - Cliffs Low Temperature Overpressure Protection >
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System -
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- Gentlemen:
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- On1 November 27,.1989, a meeting was. held with. Nuclear Regulatory Commission (NRC) j p
representatives to4 discuss our = current Low Temperature Overpressure Protection -- (LTOP)
' system.. The NRC ~ requested that we provide a description of - our system. Attached is q
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that description.
We will( comply with the controls defined in this letter while the NRC reviews a
. subsequent Technical Specification 1 submittal. These controls -are more-conservative, in-u
- terms L. of ' plant. operation,' than-the current Technical - Specifications. - We will not 0
exceed : a RCS temperature of 319 F untilCverbal concurrence -is - obtained from the NRC.
1These commitments were-agreed to at the November 27, 1989, meeting ' described above.
Very truly s,-
g, STATE OF MARYLAND
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7 I herib cer fy. that on the N U day of t hiu s.tu
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. subscriber, a Notary Public of the State of. Maryland in-.and for dF d4a 'A / -
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, personally appeared George C. Creel,' being diily sworn, Md states
't',at.he: is ' Viee President - of the Baltimore Gas and Electric Company, a corporation of-
' the State = of Maryland; that he provides the foregoing response for the purposes therein set forth;.. that the: statements made are true and correct to the best of his knowledge, infcrmation, J and ' belief; and that he was authorized to provide the response on behalf. -
yW Jof said Corporation.
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February 7,:1990
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D. A. Brune,, Esquire
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Silberg,' Esquire i
R. A.Capra, NRC:
D. G. Mcdonald,' Jr., NRC-
' W. T. Russell, NRC -
s' Ji - E. Beall, NRC '
T. Magette, DNR' f
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s ATTACHMENT- (1)
LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM DISCUSSION in 1976, the Nuclear Regulatory Commission (NRC) required commercial nuclear power plants to institute automatic and administrative controls to prevent exceeding the 10 CFR 50, Appendix G operating limits (Pressure-Temperature Limits) for the reactor
- vessel during operations at low temperature. The Unit i Low Temperature Overpressure Protection (LTOP) controls were based on Pressure-Temperatures (P-T) ' limits for the reactor vessel that were applicable for the first 10 Effective Full Power Years (EFPY) of operation (Figure 3.4-2a in the Technical Specifications). Operation has now shifted to the P-T.
limits that are applicable for 10-to-40 EFPY of operation (Figure 3.4-2b in the Technical Specifications). Using the 10-to-40 EFPY F-T limits as a basis for LTOP controls, however, would severely impact plant operation at low temperature since an. insufficient pressure band would be available for operation of reactor coolant pumps, in 1987, we reviewed the draft revision to Regulatory Guide 1.99 which addressed a change in the potential for brittle fracture of the reactor vessel at low temperatures.
We realized that our reactor vessels might be affected by the material property changes described ~ in the draft Regulatory Guide. Because of these concerns, which reduce our operating window, we requested Southwest Research Institute (SwRI) to provide 10 CFR 50, Appendix G heatup and cooldown curves for 12 EFPY The method described in the draft - Regulatory Guide was used to generate these curves.
Attachment 2
of Reference (a) is the report that SwRI provided to describe their analysis. This
' report was reviewed by BG&E while in the draft stage. Review was provided in the following areas:
neutron fluence,- material properties assessment, and licensing.
Comments were provided to SwRI in the areas of neutron fluence and material properties assessment.
These comments were satisfactorily incorporated and the analysis was accepted by BG&E in early 1989.
11eatun and Cooldown Rates -
The 12 EFPY heatup and cooldown curves are less restrictive than our 10-40 EFPY curves. The 12 EFPY curves are used to determine the heatup and cooldown rates and the LTOP 'setpoints (enable temperature and pressure) which protect the reactor. vessel against brittle fracture. The heatup and cooldown rates for the 12 EFPY curves are more conservative than the rates stated in Technical Specification 3.4.9.1.
Ilowever, because the 10-40 EFPY curves are in our Technical Specifications, those curves govern the heatup and cooldown pressures and temperatures. Operating Procedure
-OP-1 and OP-5 describe operations while in the LTOP condition. These procedures use the heatup and cooldown rates based on the 12 EFPY curves but require the operators to maintain -RCS temperature and pressure below the 10-40 EFPY curves.
Maximum Pressure and Temnerature (MPT) Enable Temnerature Setnoint The enable temperature is determined using the method given in Standard Review Plan (SRP) 5.2.2., Overpressure Protection.
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- ATTACHMENT (1)
LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM From. SRP. 5.2.2,.- the - Adjusted Reference Temperature ' (ART) at the controlling location is added to 90 F to determine the enable temperature. The difference between 0
the metal temperature.and coolant - temperature must - also be accounted for, along with
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instrument uncertainty. -From L Reference (a) of Attachment (2), for Unit I at 12 EFPY,-
0
'we have an ART at 1/4T = 219 F and an ART at 3/4T - 166 F, 1/4T is ~ the.
-. controlling location for cooldowns. and 3/4T is - the controlling location for heat-ups.
' During a t cooldown, the metal temperature lags the coolant temperature so no temperature 0
0 correction is necessary. The enable temperature for cooldowns is ' 219 F + 90 F = 309 F.
During -- a heatup, the ' 3/4T metal temperature lags the coolant temperature by 43.2 F. So 0
0 the enable temperature for a heatup is - 166 F + 90 F + 43.2 F '= 299.2 F. Since the MPT-0 enable temperature calculated for cooldowns is greater than that for a heat-u'p, 309 F is chosen as the, basis of our MPT enable setpoint. A 10 F instrument uncertainty is applied which makes the MPT enable temperature 319 F.
This process is formally documented in NEU calculation 100-MS-8906 (Reference b).
Basis For LTOP ' Pressure Setooint
-The LTOP pressure setpoint is chosen to protect against brittle fracture. Since Unit 1 has Just exceeded 10 EFPY, protecting the level of embrittlement at 12~ EFPY is sufficient to provide protection - until the NRC completes its review of our response to, Generic Letter 88-11. The.12 EFPY heatup and cooldown curves.are shown in Figures 6.1 and 6.2 of Attachment (2) to Reference (a). The analysis supporting a
these -curves represent conservative limits for brittle fracture protection.
To i
determine the. LTOP power-operated relief valve (PORV).. low pressure setpoint, the followin's procedure was used:
L 1.
The - SwRI generated -beltline pressure - data are. corrected for the static and dynamic differences between pressure at the beltline and pressure in -the pressurizer. For a cooldown, with no RCPs running, a 15 psia static head difference is : subtracted from the beltline pressure. For-. a heatup, with RCPs running, the dynamic pressure difference (52 psia)- is subtracted from the f
- beltline pressure. This yields the indicated pressurizer pressure, 2.
Loop instrument uncertainty (38.3 psi) is added to the corrected " pressurizer pressure". This final value is the LTOP setpoint (422.7 psia).
This, process is documented in NEU calculation 100-MS-8905 (Reference c) and I&C :
calculation (Reference d).
b Administrative Controls The. LTOP setpoint and enable temperatures do not protect against the full range of mass and energy addition transients. Administrative controls are required to mitigate or eliminate the limiting transients.
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. ATTACHMENT- (1) 1 E
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LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM i
1..
Mass Addition" i
Peak pressures for mass addition transients are calculated by using the flow.
1 curve superposition method as shown in Figure
- 1. The PORY flow is calculated as inertial flow using - the Darcy-Weisbach formula which requires flow area,- fluid 1"
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. density,- pressure - drop and a loss factor. The flow area was determined' by Electric Power Research Institute PORY - flow tests. Our PORVs. have an EPRI test coefficient of 0.95.
This coefficient effectively reduces PORY flow - area by SE The PORY discharges to a quench tank equipped with a disk designed to rupture at 115 - psia.
Therefore - the. PORY pressure drop conservatively assumes a constant back pressure of 115 psia. The ' PORV total-loss factor was determined by combining 1
the. equivalent loss-factors of the PORY and piping from the PORV to the i1 quench tank. ' From Figure 1,
it is evident that HPSI pump starts must be '-
eliminated or alternately HPSI flow must be th:ottled. When used, HPSI flow l
'l must be - throttled. to 350 gpm or less to prevent. exceeding the PORY capacity at the LTOP pressure setpoint.
-1 2.
Enerav Addition Transient The. limiting energy addition transient is the start of a reactor coolant pump
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. while - steam - generator secondary. temperature is greater than primary coolant 1
temperature.
To - prevent PORV ~ lifts, reactor coolant pumps are not permitted to start while -
the? reactor coolant system (RCS) is solid. With a ' bubble in the pressurizer, pump starts with a steam generator: secondary to primary delta T -of less than H
0 150 F are permitted based on thermal hydraulic transient analysis which show ' peak pressure at 10 minutes to be less than the PORY: lift setting. The reactor coolant : pump.(RCP) start - with a - bubble in the pressurizer was modeled using - a
-RETRAN four loop model simila'r to that presented in the BG&E RETRAN Top; cal report (Reference c).' Analysis specific differences - include. removing the primary heat slabs and reducing the secondary model to a ' single node for each -steam generator. A diagram of the model is given as Figure 2.
ORIGINAL SYSTEM DESIGN Reference (f) : provided. the NRC with a description of-our LTOP system. The NRC approved. this basic system ~ in References (g) ' and (h).
After a meeting held' i
November 27,1989 'with the NRC to discuss our current LTOP system, we reviewed our original-commitments for LTOP. During this review, we identified 38 commitments.
The results of this review, along with the commitment closeout packages, are available for inspection. A - discussion of how we are handling the more significant commitment I
- deviations is provided below.
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~ 'ITACHMENT (1)
A
- LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM 1.
PORV Dischgree Pininn During an LTOP transient, the PORY may relieve water. We had committed to analyzing the. PORY piping to ensure that it would not fail when water was j
relieved through it. The' analysis was not performed at. the time. After this commitment.was identified, an analysis was performed and the piping was shown to
. be able to withstand a liquid ' discharge without loss of function, 2..
ECCS ' Testina j
To -. limit the probability Jof a mass addition transient, we had committed to -
prohibit ECCS testing when the plant was in a cold, water solid condition.
Although - testing has never been performed under water solid conditions that prohibition was added to the operating procedures in 1988. The current operating procedures contain 'a caution statement which prohibits ECCS testing when in a.
water solid condition. The appropriate Surveillance Test Procedures are being
' modified to contain a similar statement.
3.
HPSI - Header Isolation Valves We had committed.to lock the llPSI header isolation valves shut when the plant was, in n' water : solid. condition to. provide defense-in-depth against a mass addition transient. Procedures were not developed to conform to this commitment.
The operating procedures have now been: changed to require that the header isolation valves be shut, with their breakers tagged out, when the plant is in a water solid condition.
4.
Comnuter Generated Alarm For additional operator information, we committed to have an alarm with setpoints below the LTOP pressure setpoint. This ' would warn the operator - that an overpressure excursion was in progress. Although this alarm has existed since
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'1977, it ' was.not designed exactly as described in our original documentation.
Only one pressure channel was hooked up to the computer alarm. There is also r
-some, question as to what was intended as the audible portion of the alarm. Work
'is currently underway to provide full alarm function, including an audible alarm, N
as described. in the original submittal. That work' will be completed prior to
' entry into MODE 3.
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- 5.
' Steaminn the Steam Generators to 220 E To prevent an energy addition transient, we committed that the RCS temperature would be close to that of the secondary side of the steam generators during a cooldown. We committed to use the steam generators to cool the plant down to 220 F before putting the RCS on shutdown cooling. We normally use the steam r
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ATTACHMENT (1) 1 LOW TEMPERATURE OVERPRESSURE PROTECTION SYJFEM generators concurrent with shutdown cooling to get the RCS temperature to less than 220 F, This minimizes the temperature - difference between the RCS and the 0
steam ' generator secondary side. It is not practical to use only the steam generators to cool the RCS at such low temperatures. We will continue to use shutdown cooling concurrent with steaming the generators ' until the RCS temperature. is' below 220 F.
Additionally, Technical Specification changes are 0
being proposed which would eliminate reactor coolant pump. starts whenever -the RCS ' W, wrature is more than 150 F cooler than the steam generator secondary
- side, i
~ 6.
HPSI Controls To prevent a mass addition-transient that could threaten the vessel integrity, we originally committed to disable the HPSI pumps in stages. One IIPSI pump would be taken - out of service at 310 F, ' the second at 220 F and the third at-0
'160 F, Constraints have been identified since 1977 which. prevented our verbatim 0
compliance - with this control. Those changes are the need to have a HPSI pump-available for boration in MODES 5 and 6 and the requirement to have a HPSI pump OPERABLE to substitute for a loss of shutdown cooling (Generic '
Letter 88-17). As a result of these - constraints, we have implemented an alternative set of HPSI pump controls. We - will rack out the breakers for two HPSI pumps' and place the third in pull-to-lock prior to the MPT enable j
setpoint. This provides assurance that the HPSI pump will not ' inadvertently inject water into the vessel, yet leaves a HPSI pump available to gerform other functions. When a HPSI pump is used at RCS temperatures s 319 F, either the flow must. be throttled, or an adequate vent in the. RCS provided. To provide j
defense-in-depth below 319 F, the HPSI loop isolation valves will~ be ' prevented from operating. automatically. This prevents n-SIAS from causing: the. loop-isolation valves to open and provides defense-in-depth' when the plant is not in a
'l water solid condition.
l SAFETY SIGNIFICANCE
' The ; Unit' 1 12-EFPY P-T limits were conservatively developed _ in accordance with the fracture toughness requirements of 10 CFR 50, Appendix G as supplemented by the ASME 111, Appendix G. The mechanical _-properties and chemical composition of i
Code Section the / reactor vessel beltline materials used in the analysis were the same as those _used to ' evaluate the Pressurized Thermal Shock ~ (PTS) concern in -January _1986. These material characteristics have been reviewed and accepted by the NRC as stated in Reference (i). The peak reactor ' vessel fluence was calculated using Discrete Ordinate Transport (DOT) calculations with a DOT. IV.3 computer code. The analysis of the reactor vessel material irra_diation surveillance specimens was used to. verify the p
- validity of the fluence calculations. The Adjusted RT values were based on the NDT conservative methodology provided in Regulatory Guide 1.99, Revision 2 (Proposed).
The 12 : EFPY curves provide a large margin of safety as required by 10 CFR 50, Appendix G, as supplemented by ASME Code Section III, Appendix G. The conservative methodology of Regulatory Guide 1.99, Revision 2 (Proposed) also provides a large margin ' of safety for the prediction of reactor vessel neutron embrittlement..
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I ATTACHMENT (1)
~ LOW TEMPERATURE OVERPRESSURE PROTECI1ON SYSTEM The LTOP ' controls have been revised to ensure compliance with - the 12 EFPY P-T limits. Administrative controls have been revised to prevent occurrence of events for which automatic protection is insufficient to prevent exceeding-the P-T limits. The' PORV low pressure setpoint has been reduced from 450 psig (464.7 psia) to 422.7 psia to~ ensure automatic protection is ' provided during solid plant conditions. A thermal hydraulic LTOP analysis - has been performed to establish the peak system pressure for the limiting mass and energy addition transients. The revised P-T limits provide conservative limits on reactor coolant system pressure to minimize the likelihood of a rapidly propagating fracture - due to pressun transients at low - temperature. The
-- revised LTOP controls provide Adequate protection assuming failure of the. most i
limiting single active component.
The - revised - LTOP controls provide adequate
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assurance that the P-T limits will not be exceeded during normal operation and anticipated operational ' occurrences in the low temperature region.
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A*ITACHMENT ' (2) '
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t REFERENCES (a)~
Letter from.GiC. Creel (BG&E), to. Document Control Desk dated October 27, 1989,.
Request for Amendment-
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NEU Calculation 100-MS-8906.
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NEU Calculation 100-MS-8905 '
(d)i. I' A.C Calculation.1-89-118-
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.(e)
BG&E RETRAN Topical Report, A-85-ll A, dated January 31, 1986
-(f).. Letter from. V. R.. Evans.,(BG&E) to D. ' K. Davis (NRC), - dated July: 21-1977, j
Reactor : Coolant. System Overpressurization t
(g) !--
Letter from 'R L. Baer (NRC) to K. R. 'Goller (NRC) dated. November.17,' 1977,
- Safety Evaluation of"the ' Low Temperature Overpressure Protection ' System l(h)
- Letter' from - R. W. Reid (NRC) to A. E, Lunduall.(BG&E) dated August 7,1978,.
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~. Issuance of' Amendments Nos. 34 and 16.-
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Letter from S. A. McNeil (NRC) to J. A. Tiernan (BG&E) dated February 4,-'1987,'
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Projected L Values-of. Material' Properties for Fracture Toughness Requirements' for ~
4 Protection.1 Against Pressurized. Thermal Shock Events.
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