ML20207F032
| ML20207F032 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 07/17/1986 |
| From: | Farrar D COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 1850K, NUDOCS 8607220469 | |
| Download: ML20207F032 (13) | |
Text
l
/e Commonwealth Edison s
l Z
72 West Adams Street, Chic 5go,!!!inois V
Address ReplyEPost Office Box 767 Chicago, lilinois 60690-0767 July 17, 1986 Mr. Harold R. Denton, Director U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC. 20555
Subject:
Braidwood Station Units 1 & 2 Proof & Review Technical Specification Comments NRC Docket 50-456 & 50-457 References (a):
April 7, 1986 A.D. Miosi letter to H.R. Denton (b):
June 10, 1986 J.A.
Stevens letter to D.L. Farrar (c):
July 1, 1986 A.D. Miosi letter to H.R. Denton
Dear Mr. Denton:
This letter provides Commonwealth Edison Companys' comments concerning three remaining issues with the proposed Technical Specifications for Braidwood Station.
These issues include the need for limiting Conditions fo'r operation at the station based on equipment for the mitigation of certain high energy line breaks, batteries for the diesel-driven auxiliary feedwater pumps, and additional seismic monitoring equipment.
In Enclosure One, we conclude that if the equipment which has been installed to detect and mitigate certain low probability breaks in high energy lines fails to operate, there is minimal potential for impacting the capability to safely shutdown the plant.
In addition we believe, based on the criteria which we understand are included in the Commission's proposed Policy Statement on the content of Technical Specifications, that these events 60 not meet the threshold of importance for which Technical Specifications should exist.
On the other hand, the batteries for the diesel-driven auxiliary feedwater pumps are clearly included, we believe, in the l
technical specifications for the pumps because of the standardized definition of " Operable / Operability".
As a result of our understanding, we verify the operability of the batteries during the monthly test of the pumps.
Other examples also similarly exist for other ancilliary equipment necessary for some function of major systems included in the Technical Specifications.
And in further support of our understanding of this definition, we point out in gt 72M; M9h6 po/
IIi
l
. Enclosure Two that previous staff positions reached for our Byron station have not included these batteries as a separate item in the Byron technical specifications.
Finally, concerning the need to install additional seismic monitoring equipment, we are providing in Enclosure Three a review which demonstrates that the appropriate regulations as well as the guidance in the Standard Review Plan have been met by the five currently installed seismic monitors.
The Braidwood design is identical with that at our Byron station, which was accepted in the Byron SER, with one exception.
An additional monitor was installed at Byron because the seismic response spectra of the river screen house differed from that of other Category I structures.
At Braidwood, the lake screen house seismic response spectra does not differ from the response spectra for the containment foundation.
One signed original and fifteen copies of this letter are being provided to facilitate your review of our comments.
Very truly yours,
.)
_ =__,
D.
L.
Farrar Director of Nuclear Licensing
/klj cc:
J. A.
Stevens L. Olshan encl.
1850K
ENCLOSURE 1 High Energy Line Break Tech Specs. For Steam Generator Blowdown and Auxiliary Steam Lines In the unlikely event of a steam generator blowdown line or auxiliary steam line break or crack, the installed temperature switches are designed to close valves which will isolate the failed lines prior to exceeding the environmental qualification test temperatures in the areas which contain safe shutdown equipment.
Redundant safety related sensing circuits and redundant isolation valves provide a highly reliable system.
The probability of total unavailability of the isolation system is small.
The system is made up of reliable components which are safety related.
Full redundancy is included in the design.
If part or all of an isolation system was temporarily out-of-service, the resulting increase in risk given an event which would prevent safe shutdown is very small such that technical specifications which require plant shutdown are not justified.
The adverse effects of an unnecessary plant shutdown and transient are more significant than the potential for an accident.
In addition, the criteria under consideration by NRC in the draft policy statement soon to be published for comment, focuses on accidents and transients addressed in Chapters 6 and 15 of the FSAR.
These HELB's do not meet the threshold and are only discussed in Chapter 3 of the FSAR.
The event in question is of itself very low probability.
Failures of seismically supported high energy piping in nuclear power plants are very rare and are generally assigned probabilities on the order of 10-4 per reactor year.
Even this probability is conservative for these events because the most probable failure, a small crack with very slow propagation, is not of concern.
The automatic isolation system was added because of the analytical difficulty in establishing that adequate time was available for operator action following a full double ended piping rupture.
However, available time for operator response increases as the break size decreases.
Environmental effects of small cracks would be mitigated by the HVAC system and would not result in unacceptable environmental conditions.
Even if a full double ended break did occur, and the isolation system was not operable, it is very unlikely that equipment would fail.
Because of the size and complexity of the required calculations and the uncertainties involved, no attempt has been made to calculate the resultant temperature which would be experienced by safe shutdown equipment.
Since the postulated blowdown is from the subcooled liquid portion of the steam generator and no significant pressurization will occur, the areas near the break will be at 2120F or less.
Because of the limitations in the break flow, locations farther from the break will experience lower temperatures.
The exact temperature at any location is a complicated function of the break size, the distance from the break, the initial temperatures, and the station HVAC system performance in the area.
Preliminary calculations were performed which give some insight into the expected effects in the Auxiliary Building The areas which contain Safety Related Equipment and are closest to the break have an existing normal qualification temperature of 1400F.
The equipment is designed for and expected to operate normally in temperatures up to this level.
A scenario of maximum break size with conservative but realistic initial temperatures and HVAC performance resulted in an environmental temperature of about 180oF at the nearest Safety Related equipment.
Because of mixing with the normal ventilation flow, the potential for significant condensation on equipment is low, although the relative humidity would be somewhat higher than normal.
The areas farther from the break will be exposed to lower temperatures and humidities.
Elevation 364'-0 is of most concern since the breaks occur on this elevation.
The temperatures would be higher and rise faster on this level than on others.
There are no significant "short circuit" paths to other elevations other than the stairways in the center of the building.
On elevation 364'-0 the safety related components can be categorized as cables, equipment in compartments, and equipment in the general areas.
Cables would not be adversely affected by temperatures up to 1800F for short periods of time.
i Important emergency equipment such as the Charging Pumps and the Safety Injection Pumps are located in subcompartments and provided with local fan coolers which are in operation whenever the pump is in operation.
There would be no need for Safety Injection in this event and the Charging Pumps, which are normally running, would be cooled by the local cooler.
Other equipment on this elevation is located in relatively open areas.
The most significant equipment is the Component Cooling Pumps and Valves, Motor Control Centers 131X1, 132X1, and 231X1, Reactor Coolant pressure transmitters, and various containment isolation valves.
Much of the equipment can be seen to be unnecessary for the event in question because the break would not result in reactor shutdown, loss of offsite power or require i
._,m-,_----,--..-..,..._.m.
_,-c...,...m.m__._._--_
containment isolation.
The RC pressure transmitters are utilized
)
only for a monitoring function and as an input for interlocks which are in effect post-LOCA.
MCC 131X1 is associated with many components.
Most significant are the 1A Charging Pump local cooler Diesel Generator 1A fuel transfer pump, the Essential Service Water i
pump 1A local coolers.
MCC 132X1 is associated with the following important components:
Essential Service Water pump 1B local coolers, and Diesel Generator 1B fuel transfer pump.
MCC 231X1 is a Unit 2 MCC with loads corresponding to the Unit 1 loads on 131X1.
With the exception of the Component Cooling system, the affected equipment is either not required for the postulated scenario or would be required only if a specific unrelated single active failure (i.e. loss of offsite power) is postulated.
This further reduces the probabilities of a safety significant problem by a large factor.
The component cooling system consists of five pumps which serve two units.
One functional pump per unit is adequate for safety requirements for non-LOCA events.
In a non-LOCA event the immediate effects of component cooling loss are limited.
Reactor Coolant Pump thermal barrier flow is lost but the seal injection system will maintain the RC pressure barrier and the system operability.
Loss of the cooling flow to the letdown heat exchangers would require isolation of the letdown system but, again, this would not pose a safety or an immediate operability problem.
If the plant were to be shut down, loss of flow to the RHR heat exchangers would prevent going to cold shutdown conditions but would not jeopardize safety in the hot shutdown condition.
Although the MCC's are not airtight, the construction will prevent high levels of infiltration.
This, in conjunction with the limitations on heat conductivity of degradable (non-metallic) parts, will increase the heat lag and prevent the actual components from reaching the postulated environmental temperatures.
Although the Component Cooling pump motors are not qualified for harsh environments, they are designed to continuously operate at normal ambient temperature.
The short transient associated with the worst postulated break which peaks at 180cP would not be expected to result in failure of the motors.
Furthermore, our planned administrative surveillance program will require a channel calibration and analog channel operational test each refueling outage.
For all these reasons, the equipment used to monitor HELB temperature should not be included in the Tech Specs.
~_
ENCLOSURE 2 Individual Tech Specs for Diesel-Driven Auxiliary Feedwater Pump Batteries We have concluded that a technical specification for the batteries that support operation of the diesel-driven auxiliary feedwater pump are not necessary.
This position is based on the Technical Specification definition of Operable-operability, a highly reliable design, and previous staff positions.
By generic letter dated April 10, 1980, the Staff indicates that support systems (such as electrical power or water systems) are encompassed by the definition of OPERABILITY as set forth in section 1 of Technical Specifications.
This position was reaffirmed in a memorandum covering Technical Specification Operability requirements from D.M. Crutchfield of NRR dated July 8, 1985.
In addition, the staff's position is inconsistent with similar support systems associated with other Technical Specifications.
Some examples are:
a.
The air-start systems (or lube oil or' jacket water) for the station diesel generators.
b.
The hydraulic accumulators associated with the MSIV's.
c.
There are virtually no LCOs or surveillances specifically dedicated to pump motor integrity or lube oil systems.
d.
There is no specific surveillance requirement to ensure the proper operation of the overspeed trip, cooling water or other support systems for the auxiliary feedwater diesel.
The proposal to impose a new Technical Specification ~
requirement is unnecessary.
The diesel engine starting systems consist of two full capacity battery banks and associated chargers.
The chargers are powered from a Class lE bus.
Both banks of batteries have surveillances performed on a daily, weekly, quarterly, and refueling outage basis.
In addition, the draw-down voltage of each bank is checked on a staggered basis during the monthly pump run.
Presently, the chargers are connected during the performance run.
If required a procedure change can be implemented which would disconnect the charger from the battery during the performance run.
The Staff has provided Commonwealth Edison with a draft IEEE standard applicable for nickel-cadmium batteries in a constant duty application for assessment of the adequacy of the present level of surveillances.
The stations' surveillances are in compliance with the applicable recommendations of the draft IEEE standard with the exception of designating a pilot cell.
The imposition of a Technical Specification requirement for these batteries would do nothing to enhance their availability and/or reliability.
Adequate administrative control already exists.
This is evidenced by:
1.
Surveillance schedules are controlled in accordance with an approved administrative procedure.
2.
The surveillances associated with the subject batteries are performed using approved station procedures.
Any changes in level or frequency of surveillance must be evaluated using the criteria of 10CFR50.59 and approved by the On-Site Review Committee.
3.
In response to FSAR Question 040.68, Commonwealth Edison committed to performing surveillance and maintaining the subject batteries.
In a reliability analysis performed by Torrey Pines Technology on the Byron /Braidwood Auxiliary Feedwater System, the specific Technical Specification limitations and surveillances were presented in Sections 3.6 and 3.7.
These limitations and surveillances did not include a specific LCO or surveillance for the engine starting system.
This report and the proposed auxiliary feedwater system Technical Specifications were specifically found to be acceptable in Section 10.4.9 of NUREG 0876, the Byron SER (also applicable to Braidwood).
Data for this study was extracted from NUREG 0611, which included the only other plant using a diesel-driven auxiliary feedwater pump.
In NUREG 0611, no failures were attributed as a result of a degraded battery.
The station battery LCO and associated surveillances are written to ensure the operability of lead-calcium 125 VDC ESF batteries which are employed in a constant-duty capacity.
The expansion of this-LCO to include a 24VDC nickel-cadmium battery used in an engine starting application is inappropriate.
The surveillances applicable to the station batteries are not applicable to demonstrate the ability of a battery in an engine starting application to fulfill its design function.
The Staff's position constitutes a backfit under 10CFRSO.109 in that:
a.
It constitutes a change from a previously applicable Staff position.
In this case the previous Staff positions are reflected in:
1.
NUREG 0876 2.
Generic Letter dated 4/10/86 3.
Crutchfield memo of 7/8/85 4.
Byron Units 1 and 2 Technical Specifications b.
It was identified in writing to the licensee:
1.
After O.L.
issuance, in the case of Byron.
2.
After O.L.
application, in the case of Braidwood 3.
After design approval for the duplicate plants at Byron and Braidwood (Appendix N).
4.
After CP issuance for both plants.
c.
The imposition of the proposed Technical Specification requirement will not result in a i
substantial increase in the overall protection of the health and safety of the public because:
1.
The Byron /Braidwood AFW systems have been found to be in the acceptable range for unavailability by the Staff in NUREGs 0611 and 0876, and 2.
The imposition of the proposed Technical Specification would not impact the availability in that the present level of surveillance exhibits a high degree of compliance with available industry guidelines.
..,, ~. -.
. _ _,..... -. - _ _ -, _. ~. -. _.. -.. _ -
ENCLOSURE 3 The seismic instrumentation was reviewed in the Byron Safety Evaluation Report, Section 3.7.4 with the following conclusion:
I "The type, number, location, and utilization of strong motion accelerographs to record seismic events and to provide data on the frequency, amplitude, and phase relationship of the seismic response of the containment structure comply with Regulatory Guide 1.12."
The Standard Review Plan, Section 3.7.4.II.1 states, in part, "The seismic instrumentation program is considered to be acceptable if it is in accordance with Regulatory Guide 1.12....".
Therefore the Byron seismic instrumentation is in compliance with Regulatory Guide 1.12 and the Standard Review Plan.
The Braidwood instrumentation configuration is identical to the Byron configuration, with one exception.
Byron has an independent Category I structure, the River Screen House, which requires a separate response spectrum recorder because its response is different than the containment structure.
Braidwood has no such seismically independent Category I structure.
This means that Byron has five sensors to monitor the containment structure, and one sensor to monitor an independent structure.
Braidwood has the same j
five sensors for its containment structure, and no seismically independent Category I structure, therefore a sensor for an independent Category I structure is not required.
We have addressed the subject of seismic response at the i
Braidwood Lake Screen House in the FSAR in response to NRC Questions 362.11 and 130.56.
Question 362.11 was a request to determine amplification of the rock spectrum due to the soil between the rock and the foundation of all Category I Structures, and Question 130.56 1
requested a description of the method of analysis used to account for the soil-structure interaction at the Lake Screen House.
Our response to these questions showed that the soil between the Lake Screen House foundation and rock is of adequate stiffness, so that l
no amplification of the rock spectrum will occur.
The Lake Screen House rests on 10 feet of hard glacial till.
The shear wave velocity of the till material is 2400 ft/sec.
The shear wave velocity of the underlying rock is 3200 ft/sec.
Because of the high soil column frequency (Vs/4H = 60 Hz) and the low velocity contrast (3200/2400 = 1.33) between the rock and the soil medium, there will be no appreciable amplification of motion between the rock and the i
top of the till in the critical frequency range of 1 to 20 Hz.
l Thus, the seismic response at the foundation of the Lake Screen House will be the same as at the containment foundation, which rests on rock.
1 i
,,,,_.---...n.
--._e-.,--------.,,-,,
,..--,e
It is acknowledged that the Regulatory Guide 1.12 requires six or more response functions to be performed by the instrumentation, but a requirement for six discrete sensors is not contained in Regulatory Guide 1.12, the Standard Review Plan, ANSI /ANS Standard 2.2-1978, or ANSI Standard N18.5-1974.
The six response functions are provided by the present design.
A detailed review of the Braidwood seismic instrumentation is provided on Attachment A,
" Comparison of Braidwood Seismic Instrumentation with RG 1.12 Requirements" and Attachment B,
" Comparison of Braidwood Active Seismic Instrumentation with SRP Table 3.7.4-1."
We have concluded by reviewing the existing regulations and the Byron and Braidwood Safety Evaluations, that the current design at Braidwood is acceptable.
1850K l
l l
l
Enclosura 3 Attachment A COMPARISON OF BRAIDWD00 SEISMIC INSTRUMENTATION WITH RG 1.12 REQUIREMENTS REG. GUIDE 1.12/ ANSI N18.5 BRAIDWOOD SEISMIC INSTRUMENTATION PROVIDED C. REGULATORY POSITION Earthquake instrumentation specified in ANSI N18.5, " Earth-quake Instrumentation Criteria for Nuclear Power Plants," is acceptable to the Regulatory staff for satisfying the seismic instrumentation requirements indicated in Paragraph VI (a) (3) of Appendix A to 10 CFR Part 100 for assuring the safety of nuclear power plants, subject to the following:
1.
The instrumentation called for in Section 4.1 of the Standard should be applied to nuclear power plants with a Safe Shutdown Earthquake maximum foundation acceleration of less than 0.3 g as supplemented by the following:
a.
Instead of the locations specified in Section 4.1.2 of the Standard, one triaxial peak accelerograph should be provided at one location of each of the following:
(1) A selected location on On accumulator tank at Elevation 426'-0" in reactor equipment.
the containment building.
(2) A selected location on On safety injection piping at Elevation 421'-0" the reactor piping.
in the containment building.
(3) The most pertinent location on one of the following outside of the containment structure:
(a) Seismic Category I equipment (b) Seismic Category On essential service water return piping at I piping.
Elevation 346'-0" in the Auxiliary building.
~
4.1 Safe Shutdown Earthquake Maximum Ground Acceleration of Less than 0.2g.
ANSI N18.5 4.1.1 One Triaxial Time-History Accelerograph shall be provided at one location of each of the following:
(a) " Free fleid." See note to (b).
At site coordinates 39+00E, 41+005
Enclosura 3 REG. GUIDE 1.12/ ANSI N18.5 BRAIDWOOD SEISMIC INSTRUMENTATION PROVIDED (b) Containment foundation.
At Elevation 377'-0", Azimuth 1450 This ANSI Note: If soil-structure interaction sensor provides input to both the Time-History N18.5 is negligible, a single instrument Accelerograph (THA) and the Response Spectrum may be located on the " free field" Recorder (RSR).
or the containment foundation.
(c) Containment Structure or On containment shell wall at Elevation 502'-0",
reactor building.
Azimuth 1450 b.
One triaxial response-spectrum Base slab of containment building at recorder capable of measuring both Elevation 377'-0".
This sensor is the horizontal motions and the vertical same as 4.1.1b above and provides input motion and capable of providing signals to both the THA and RSR.
for inmediate control room indication should be provided at the containment foundation.
c.
One triaxial response-spectrum recorder capable of measuring both horizontal motions and the vertical motion should be provided at one location of each of the following:
(1) A selected location on the On containment refueling floor at Elevation 426'-0".
reactor eqaipment or piping supports.
This sensor measures the input to the accumulator tanks. The TH Response is recorded for this location and then used as input to the RSR.
(2) The most pertinent location on one of the following outside of the containment structure:
(a) A Seismic Category I At Elevation 426'-0", Counting Room floor equipment support or in the auxiliary building.
appropriate floor location 4
(b) A Seismic Category I piping support or appropriate floor location.
4 (3) At the foundation of an independent None provided, at Braidwood there is no Seismic Category I structure where the Category I structure with foundation response is different from that of the response different from the containment.
reactor containment structure.
8112b
]
Attachment B J
Comparison of Braidwood Active seismic Instrumentation with SRP Table 3.7.4-1 Standard Review Plan Byron /Braidwood Location / Type Braidwood Location FSAR Reference I.
Free Field Site coordinates 3.7.4.2.2.1 Time-History Accelerograph 39+00E, 41+00S (THA)
II.
Inside containment Basement - THA Elevation 377' (dual 3.7.4.2.2.2 sensor)
Basement-Response Elevation 377' 3.7.4.2.4a Spectrum Recorder (RSR)
At Elevation - THA Elevation 502' 3.7.4.2.2.3 Reactor Equipment Elevation 426' 3.7.4.2.2.4 Supports or Reactor Containment refueling Piping Supports-RSR floor III. Outside containment Cat. I Structure - RSR Not required Cat. I Equipment Elevation 426' 3.7.4.2.4b supports or Auxiliary Building Cat. I Piping Supports - RSR l
1 4
l
.w
, ~ _.,, _ _ _ - _ _. _. _ -. _
~, - - -.. < -. - -
- _ - _ _ -