ML20040C832
| ML20040C832 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 01/26/1982 |
| From: | Tramm J, Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8201290288 | |
| Download: ML20040C832 (13) | |
Text
. _ -
IN Commonwealth Edison
.I e
-- ) One First NLtionil Pitza, Chictgo 1linois C "/ Address Reply to: Post Office Box 767 Chicago, Illinois 60690 i
Janua ry 26, 1982' N
t Mr. Harold R'.
Denton, Director Of fice of Nuclear Reactor Regulation 8
i.
U.S.
Nuclea r Regulato ry Commission
/fp'O s.0J q $ [4// gS/tfgS Washington, DC 20555 j
Subject:
Byron Station Units 1 and 2 Ah Braidwood Station Units 1 and %
Advance FSAR Information N
=
NRC Docket Nos. 50-454/455/456/4 T
Dear Mr. Denton:
This is to provide advance copies of information which will be included in the Byron /Braidwood FSAR in the next amendment.
Attachment A to this letter lists the information enclosed.
One (1.) signed original'and-fifty-nine (59) copies of this letter ~are provided.
F1 f teen (15) copies 'o f the enclosures are included for your review and approval.
Please. address further questions-to this of fice.
Very truly 'yours, S8
'[~ Nuclea r Licensing Administrator T.R.
Tramm i
Pressurized Water Reactors i
Attachment j.
3 00I 3129N 3
/ l I
4 8201290280 454 PDR ADOCK 0 PDR l
.A
ATTACIDiENT A List of Enclosed Information I.
FSAR Question Responses NEW:
010.38 REVISED:
022.15 421.19 e
a
B/B-FSAR QUESTION 010.38 "Your response to 0010.15 does not analyze or evaluate the prottetive features provided safety-related equipment assuming internal missiles are generated outside of containment by failures of equipment such as valves, instrument wells, pump impellers, drive couplings and fan blades.
You state that protection is achieved by remote location or physical separation.
Provide an analysis and an evaluation of I>w those protective measures are achieved for a typical safety-related system.
The auxiliary feedwater system is considered a suitable example.
The analysis should cover the entire system including the diesel and motor driven pumps, routing in the auxiliary building and pipe tunnel, junction with its respective tempering feedwater line, and termination at the primary containment.
Equipment and pipe routing drawings should illustrate the protection afforded by spacing and separation from adjacent high or moderate energy systems and potential missile sources listed above.
The evaluation of this typical system should verify that no damage to safety-related equipment will re'sult which would prevent use of the equipment necessary to reach a safe shutdown."
RESPONSE
Safety-related systems in the auxiliary building are protected against damage from internally generated missiles by the separation, redundancy, and quality standards applied to the design of the Byron /Braidwood Stations.
As an example of the approach used, the auxiliary feedwater system has been analyzed in detail and is described here.
The auxiliary feedwater system consists of the motor and diesel driven auxiliary feedwater pumps, associated intake and discharge piping, piping in the auxiliary building and auxiliary feedwater tunnel, and the system valves and instru-mentation.
The auxiliary feedwater piping exits the auxiliary feedwater tunnel and enters the main steam tunnel where it joins the main feedwater piping.
On the pump suction side of the nystem, connections are nade to the condensate i
storage tank as a primary source of feedwater and the essential service water system as a backup source of water.
The components which are postulated to fail resulting in missiles are pumps, pump drivers, valves, and instrument wells.
None of these components are actually considered Q10.38-1 f
B/B-FSAR as a potential cause of missiles for the reasons explained in the responses to Questions 212.1, 212.12, 212.15, and 212.16.
However, the plant design incorporates additional mitigating features.
Missiles will be postulated to demonstrate the additional margin in the design.
The following events will be postulated to create missiles:
a.
Pump impeller failure b.
Pump driver (motor or engine) failure c.
Valve failure (valve stem ejection) d.
Instrument well failure.
A fracture of the pump impeller could result in ejection of fragments.
These fragments are not expected to penetrate the pump casing.
However, if penetration of the casing is also postulated, the fragments would be stopped by walls of the auxiliary feedwater pump rooms.
These rooms enclose both pumps with a 12-inch concrete wall separating the redundant pumps.
Each room contains only piping and equipment for one auxiliary feedwater loop with the exception of one short length of loop A piping which travels for a short distance through one corner of pump room B.
This pipe is routed against the ceiling and upper wall of the compartment.
The minimum distance between this line and the "B" pump or a "B" valve is 20 and 15 feet, respectively.
Therefore, failure of a pump cannot affect the redundant train.
In addition to the auxiliary feedwater system, other systems are located in these rooms.
The essential service water system and the condensate system extend into these areas to supply water to the pumps.
D& mage to one of these lines could conceivably impair delivery of water to both trains.
However, the service water and condensate piping are separated and in fact enter from opposite sides of the rooms.
Check valves in conjunction with the normally open condensate system valves and the normally closed essential service water valves insure that feedwater will be available in the event of damage to either the service water or condensate system.
Fire protection, nonessential service water, and instrument air piping also pass through these areas.
Damage to any of these systems will not impair safe shutdown of the plant.
The Byron /Braidwood plants have one motor-driven auxiliary feedwater pump and one diesel-driven auxiliary feedwater pump per unit.
No missiles are expected to result from failure of the motors or diesels.
A fragmented rotor will be contained by the stator of the electric motor.
Parts ejected following an internal failure of the diesel engine Q10.38-2
6 e
B/B-FSAR would be contained by the engine crankcase.
In the unlikely event of fragments penetrating the stator or the crankcase, damage will be limited to the room enclosing the diesel.
The loop "A" pipe mentioned above'is. located high, such that-a fragment from.the diesel would have to exit at a high angle.
This is not considered credible.
The protec-tion afforded by the rooms and the separation of systems is described above.
To address the potential of damage from valve missiles, the Unit 2 valves will be described.
There are no significant differences in the Unit 1 valves.
Gate valves are used at the junction of the condensate system (Valves 2AF002A/B) and the essential service water system (Valves 2AF006A/B) with the auxiliary feedwater system.
These valves are in the low pressure (pump suction) portion of the system and, therefore, no missiles are expected.
Additionally, thete valves are located in the auxiliary feedwater pump rooms and thus maintain the separation of the redundant systems.
Gate valves are also installed on the outlet side of the pumps.
Valve 2AF004A is just outside of pump room A and separated from all parts of the B loop by at least a 12 inch concrete wall.
Valve 2AF004B is in pump room B and is on the opposite side of the diesel from the loop A pipe-mentioned above.
Both valves are oriented such that the stem is vertical and no safety system is directly above the valve.
The auxiliary feedwater lines are routed across the auxiliary building after exiting the pump rooms.
The lines are not adjacent to any high energy or safety-related lines and contain no valves or instrument wells until the pipes (now eight; two redundant loops per each steam generator) turn down at column row 26 just prior to entering the auxiliary feedwater tunnel.
At this point is a bank of globe valves (2AF005A-H) which are used as control valves for the auxiliary feedwater flow.
These valves are mounted in a vertical pipe run.
The stems are oriented such that they do not point at any part of the auxiliary feedwater or other safety-related systems.
The lines and valves are arranged at this point such that redundant lines (to the same steam generator) are separated by at least two other lin2s.
The auxiliary feedwater lines then travel the length of the auxiliary feedwater tunnel.
There are no other valves or instruments in the line except for the globe isolation valves (Valve 2AF013A-H) which are located in the tunnel just before the lines penetrate the main steam tunnel and 010.38-3
B/B-FSAR join the main feedwater line.
These valves are all oriented such that the valve stems are vertical and no safety-related systems are routed above these valves.
Check valves in the auxiliary feedwater system are not considered as potential missiles because there is no credible failure of a check valve which would create a missile.
The only instrument well in the high pressure portion of the s; stem is the temperature sensor at the exit of the pump.
Instrument wells are not expected to fail in a manner resulting in missiles because this attachment is in a Category I Safety Class C pipe and meets code standards.
The well itself is relatively small and would not be expected to damage pipe and equipment.
Even if a missile is postulated, the instrument wells are in the pump rooms close to the pumps and could not affect the redundant system.
The investigation has established that, even with unrealis-tically conservative postulation of missiles, the auxiliary feedwater system will not be susceptible to common mode failures and will not pose a danger to other safety-related systems because of internally generated missiles.
9 e
Q10.38-4
9/B PSAR
~
QUESTION 022.15 "In accordance with the NRC request to all applicants contained in Task Action Plan A-2 of NUREG-0609, " Asymmetric Blowdown Loads on PWR Primary Systems," provide the following information for the reactor coolant system break that results in the peak differential pressure loads within the reactor cavity:
Peak and transient loading on the reactor pressure a.
vessel that includes forces and moments separated into their X,Y,Z components.
b.
Provide projected areas used to calculate these loads and the location of these areas shown on detailed plan and elevation drawings.
" NOTE:
This is a reiteration of the request made in previous question number 0022.3.j."
RESPONSE
Th[s item is answered in the r'esponse to Question 110.62 a.
and the revision to Section 3.9 of the FSAR.
b.
The drawings requested were provided under separate cover in response to Question 22.16.
t In addition, in letter WCO-256-81, L.
J.
Kripps to R.
A.
Victor, " Byron Reactor Cavity and Inspection Cavity Confir-matory Subcompartment Analysis Results," dated December 9, 1981, confirmatory dif ferential pressure loads associated with the reactor cavity were presented.
The Applicant was requested to review these results and evaluate the structural adequacy of the reactor cavity.
This evaluation has been completed and it has been determined that the existing design of the reactor cavity shield wall-is adequate to accept the pressure loadings presented in
.the subject letter. The loading combinations used in this r
evaluation were the same as was used in the original design.
I 4
Q22.15-1
B/B-FSAR
RESPONSE
FSAR Appendix B has been reviewed to determine the level of conformance with Regulatory Guide 1.94.
This regulatory guide endorses ANSI Standard N45.2.5-1974 " Supplementary Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants" as the NRC position relative to the accepted industry standard.
For structural steel, the Applicant complies with the requirements of ANSI N45.2.5-1974.
For structural. concrete, the following deviations exist and corresponding justification is provided.
B.1.2.4 Fly Ash In Process Testing ANSI N45.2.5 Table B requires in-process testing of fly ash per ASTM C618 with a frequency of every 200 tons.
Fly ash was not used in Braidwood Station concrete.
It was used en a limited basis at Byron Station.
In-process testing at Byron Stat. ion was performed by the fly ash supplier and by the Applicant's testing laboratory.
Fly ash was tested and certified-by the supplier for conformance with ASTM C618 at a frequency of every 2000 tons.
The following tests were performed every 2000 tons by the supplier:
1.
Loss on ignition 2.
3 3.
Amount retained on No. 325 sieve 4.
Sum of the oxides (SiO2 + ^1 02 3 + Fe 0 I i
23 5.
Moisture content 6.
Pozzolanic activity with lime 8.
Water requirement 9.
Soundness 10.
The Applicant's testing laboratory tested the fly ash at a frequency of every 200 tons for the following tests:
1.
Loss on ignition 2.
Sulfur trioxide (S0 )3 3.
Amount retained on No. 325 sieve.
Q421.19-2
B/B-FSAR In-process concrete control testing by the Applicant's testing laboratory at a frequency of 200 tons include chemical and physical tests for which correlation with concrete properties has been established.
Any test result that cannot be correlated with concrete properties serves no useful concrete quality control function.
Prequali fication tests were performed for every source of fly och for full compliance with ASTM C618.
Once fly ash from a power plant using a specific coal is qualified by ASTM C618, testing for loss on ignition, sulfur trioxide and the amount retained on No. 325 sieve performed every 200 tons are sufficient to ensure the uniformity of the fly ash.
Uniforming of fly ash can be correlated to concrete quality.
When fly ash was used, it was added in the proportion of 20% by weight of cement.
Fly ash was not used as a substitute for cement.
B.l.2.6 Water and Ice, Chloride Ion Content ANSI N45.2.5 has no requirement for the chloride ion content in water and ice.
ASME Boiler and Pressure Vessel Code,Section III, Division 2, subparagraph CC-2223.1 limits the chloride ion content in water to 250 ppm.
FSAR Subsection B.l.2.6 states that the maximum ion content
- did not exceed 500 ppm.
At Byron Station, the maximum chloride ion content in the mixing water does not exceed 250 ppm.
At Braidwood Station, the maximum chloride ion content in the mixing water did not exceed 347 ppm with an average content of 300 ppm.
A chloride ion content of 500 ppm is a conservative limit when compared with the limits allowed in ACI 201.2R-77 and in the proposed revision to ACI 301-72.
Limiting the chloride content in water is an indirect and easy method to limit the total soluble chloride content in the concrete.
In ACI 201.2R-77, it is stated that some forms of chloride are readily soluble cnd hence, are likely to induce corrosion in the reinforcement.
Other chlorides are not likely to induce corrosion.
- However, the test for soluble chloride is time consuming and difficult to control.
ACI 201 committee recommends testing for total chlorides and, when less than recommended maximum, states that the test for soluble chlorides is not required.
In Section 4.5.4 of ACI 201.2R-77 and in the proposed revision of ACI 301-72 (see Cvncnete International, February 1981 Issue, Table 3.4.4 on Page 55), the maximum chloride content in concrete is limited in terms of cement content, concrete exposure and type of construction.
The average total chloride content per 0421.19-3
~
B/B-FSAR cubic yard of concrete at Braidwood Station exceeds Byron Station.
The total chloride in water, cement and admixtures at Braidwood Station equals 0.025% of the weight of cement.
Of this, approx-imately 59% is provided by the water, 28% by the cement, and 12% by the admixture.
When the limits for soluble chloride in ACI 201 and the, proposed revision to ACI 301 are compared with 0.025%, it shows that this content is 2.4 times less than that allowed for prestressed concrete, four times less than that allowed for reinforced concrete in a moist environment exposed to chloride, and six times less than that allowed for reinforced concrete in a moist environment but not exposed to chloride, respectively.
At Braidwood Station, the prestressing steel is not in contact with the concrete.
Furthermore, whatever water is in contact with reinforced concrete is neither sea water nor the brackish water present on bridge decks and highways due to winter de-icing salt.
Therefore, the chlori?? induced corrosion of embedded metals in this concrete is highly unlikely.
B.1.3.3 Adjustment of Design Mixures and B.l.13 Evaluation and Acceptance of Concrete Compression Results ANSI N45.2.5 Table A requires compliance with ACI 211.1-70 and ACI 214-65 as given in the list of reference documents.
The Applicant has been requested to provide a reference which contains the two equations used and relate those equations to those contained in ACI 214.
The equations presented in FSAR Subsection B.l.3.3, for the adjustment of design mixes are the same as those in ACI 318-77 Commentary Section 4.3.1.
When the proper values of the statistical parameter t, corre-sponding to the probable frequencies in ACI 318-77 Section 4.7, are used in equations (4-la) and (4-lc) of ACI 214-77 or in Equation 7 of ACI 214-65, the two equations in Subsection B.l.3.3 are obtained.
The values given in ACI 318-77 Section 4.3.1 for the required strength, are the results obtained from equations (4-la) and (4-lc) of AC.1 214-77 when the higher of the standard deviation values are used.
B.1.10.e.: and Table B.l-3 Hot Weather Concreting ANSI N45.2.5 Section 4.5.2 requires adherence to specified requirements for hot weather concreting practice as given in ACI 305.
Q421.19-4
B/B-FSAR 4
ACI 305-72 Section 2.2.1 states that:
"For.the more massive types of heavy construction, i.e.,
those whose dimensions are'such that significant heat is generated through hydration of cement, a temperature of 60* F (16' C) or even lower would be desirable). "
FSAR Table B.1-3 allows maximum concrete temperatures up to 70* F when air temperature-is above 45" F and up to 75* F when air temperature is below 45* F.
The recommendation in ACI 305-72 applies to more massive structures than those found at Byron and Eraidwood Stations.
The containment mat foundation is the most masaive concrete element and it is much less massive than conu
'a placements for dams.
In addition, midwest hot weather. concreting conditions are mild when compared with hot weather conditions in southern regions for which the recommendations in ACI 305 were intended.
ACI 305-77 Section 2.2.2 does not contain specific concrete temperature limits, but states that:
"It is impractical to recommend a maximum limiting temperature beceuse circumstances vary widely.
Accordingly, the committee can only point out the effects of higher temperatures in concrete and advise that at some temperature, probably between 75* and 100* F there is a limit that will be found to be most favorable for best results in each hot weather operation, and such a limit should be determined for the work."
The limits in FSAR Table B.1-3 are determined to be conservative for the construction of nuclear power plants.
Table B.1-5 Fresh Concrete Testing Table B of N45.2.5 requires that the first batch produced overy day be tested for slump, air content, and temperature.
FSAR Table B.1-5 requires that the first batch of concrete used in the containment is tested for slump, air content, and temperature.
For other safety-related structures, first batch testing is not required.
Testing the first batch is intended to control overnight variations in the moisture content of aggregate, variations in the concrete materials and errors in the concrete mix proportions.
Since the batch _ plant bins and-silos are usually kept full during i
concrete production, the materials used in the next day first batch are the materials already in the plant from Q421.19-5
B/B-FSAR the preceding day of production.
Segregation, contamination, and degradation in properties of the aggregate used in the first batch of the next day are not different from those during the previous day.
Therefore, testing of the first batch of concrete will not be of any significance in controlling the quality of concrete.
Experience has shown that some variations in slump, air content, and temperature may occur several batches after production is Jtarted.
These variations are related to material transition from the materials lef t in the batch plant bins and silos overnight to those materials loaded after overnight materials are used in concrete productionm.
B l.18 and Table B.1-4, In-Process Concrete Compressive Testing ANSI N45.2.5 Table B requires that tyo cylinders for 28-day strength tests be taken every 100 yd for each class of concrete.
FSAR Table B.1-4 requires six standard cylinders for compresgive testing be prepared from concrete samples taken every 150 yd of concrete placed in Category I structures other than the containment.
Two cylinders each are tested for compressive strength at 7, 28, and 91 days.
Concrete acceptance is based on the 91 day result, however, the 7 and 28 day results were used to monitor the compressive strength development during concrete production.
Concrete testing frequency for the con-tainment conforms to the ANSI.
ACI 349-76, " Code Requirements for Safety-Related Concrete Structures," establisheg a compressive strength test frequency of one for every 150 yd of concrete placed for safety-related structures other than the containment.
Section 4.3.1 of ACI 349 allows an increase in num er u
yards representative 3
of a single test by 50 yd for each 100 psi lower than a standard deviation of 600 psi.
Table CC-5200-1 of the Summer 1981 Addenda of the ASME Boiler and Pressure Vessel Codg,if the average Section III, Division 2, allows a testing frequency of every 200 yd strength of at least the latest 30 consecutive compressive streng th tests exceed the specified streng th f by an amount expressed as:
c f
- f'c + 1*419 If'c/8.69).
cr At Byron /Braidwood Stations, the average compressive strength consistently exceed this f f r all the concrete placed.
cr B.l.16 Curing and Protec' tion Regulatory Guide 1.94 references Regulatory Guide 1.55 " Concrete Placement in Category I Structures" which endorses the use of ACI 301.
Q421.19-6
'B/B-FSAR FSAR Subsection B.l.16 items (a) and (b) take exception to the portion of ACI 301-72 Section 12.2.2 requirement that reads
" Moisture loss from surfaces placed against wooden forms or metal forms exposed to heating by the sun shall be minimized by keeping the forms wet until they can be safely removed."
The practice of wetting the forms is primarily intended for hot and dry weather conditions typical of arid regions and espacially for thin members on which wooden forms can easily desiccate.
The plastic impregnated plywood forms.used in the Byron and Braidwood Stations reduce the moisture loss to minimum regardless of being exposed to heating by the sun.
Also, the midwest summers are humid and sun radiation is not as intense as in arid regions for which the provisions in ACI 301 were intended.
Furthermore, thin concrete sections exposed to a hot and dry environment do not exist in concrete structures for nuclear power stations.
0421.19-7
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