ML19350A078
| ML19350A078 | |
| Person / Time | |
|---|---|
| Site: | Bellefonte  |
| Issue date: | 11/20/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Parris G TENNESSEE VALLEY AUTHORITY |
| References | |
| NUDOCS 8012010510 | |
| Download: ML19350A078 (11) | |
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NUCLEAR REGULATORY COMMISSION
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'. V.,o NOV 2 01350 Docket Nos.: 50-438 3
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Mr. H. G. Parris
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Dear Mr. Parris:
Subject:
Bellefonte Nuclear Plant Unit Nos.1 and 2 -
Request for Additional Information As a result of our review of your application for operating licenses for the Bellefonte Nuclear Plant, we find that we need additional information in the area' of auxiliary systems.
The specific information required is listed in E. closure 1.
Please inform us of the date when this requested additional information will be available for our review.
Please contact us if you desire any discussion or clarification of the information requested.
Sincerely, Ed? 1co Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosures:
1.
Request for Additional Information 2.
Ltr to all OL Applicants fm O. F. Ross dtd April 24, 1980 ces w/encls.: See next page f
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4 Mr.1 M. 'G. Parris Manager cf Pcwer;-
Tennessee Valley Authority 500A Chestnut Street, Tower II-
-Crattanooga, Tennessee-'37401 cc:
Herbert S. Sanger, Jr... Esq.
General Counsel Tennessee Valley Authority e
400 atennerce Avenue, Ells 33 Knoxville,, Tennessee 37902 Mr. E. G. Beasley l
Tennessee Valley Authority 4C0 Co.tnerce Avenue, 249A HBS Knoxville, Tennessee 37902 Mr..D. Terrill Licensing Engineer Tennessee Valley Authority.
400 Chestnut Street Tower - II Chattanooga, Tennessee 37401 Mr. R. A. Wallin
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Babcock & Wilcox Company P. O. Bcx 1260 1-Lynchc~ urg, Virginia 24505
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Mr. Rocert 8. Borsum 3abcocx 3.Wilcox Company Suite 420 7735 Old Georgetown Road Bethesda, Maryland 20014 Mr. J. F. Cox Tennessee Valley Authority i
400 Commerce Avenue, W10C131C
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Knoxville, Tennessee 37902 Resident Inspector I_
Bellefonte NPS
-c/o U. S. Nuclear Regulatory Commission P. 0. Box 477 Hollywood, Alabama' 35752 I
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ENCLOSURE 1
_.a REQUEST FOR ADDITIONAL INFORMATION.
Bellefonte Nuclear Plants, Units 1.and 2 Docket Nos. 50-438 and 50-439
'410.0
-Auxiliary Systems Branch 410.7 Expand Section 3.2 " Classification of Struct:..es, Components and Systems" (3.2) to include fuel handling equipment, fuel storage racks and spent fuel pool liner. ~ This list (in confonnance with CFR 10.50 App. B) is to include the seismic category, quality group and any other pertinent identification.
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Discuss the design for structures housing safety-related equipment with (3.4.1) s respect to the method for sealing all penetrations, doors, equipment hatches, or other openings located below the PMF level.
410.9 Your FSAR does not postulate potential internally generated (3.5.1.1)
- lissile scurces such as valve bonnets, instrument wells and pump impellers outside of containment.
Provide the results of an analysis and evaluation for all areas of the plant housing safety-relate'd equipment assuning potential intern 6 missiles generated from such sources as mentioned above.
This evaluation should verify that damage to safety-related ecuipment will r.ot result in preventing a safe shutdown.
410.10 Provide a tabulation of all safety-related components which are located (3.5.2) outdoors and describe the protectien to be afforded to these components y
to prevent their being damaged by tornado generated missiles. Include in i
this tabulation all HVAC system air intakes and exhausts.' The FSAR figures i
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of plant arrangements do not allow the staff sufficient detail to determine the protection afforded the air intake and exhausts. Provide drawings of sufficient scale which will provbe adequate details to make this finding.
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410.11 Regarding Vertical Tornado Generated Missiles, provide the following:
(3.5.2) a) Describ_e how plant safety is assured assuming multiple vertical tornado generated missiles striking the "A" main steam valve vent room's roof which is not missile protected.
b) Discuss the protection provided against vertical tornado generated missiles for spent fuel and safety-related equipment in the auxiliary building and other buildings which contain blow-out panels for limiting, the pressure differential on the roof (s) due to tornado winds.
410. 12 Your response to Question 10.1 concerning high energy line breaks is not (3.6.1) complete, therefore, we need the following additional infm nation:
a) Provide a complete set of legible piping area drawings marked to show
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the routing of all safety-related systems listed in FSAR Table 3.6.1-1.
These drawings should identify postulated break locations for high energy lines and should show the relative position of safety-related equipment and structures to these high energy lines.
In addition, provide a table which identifies the means of protection (f.e. pipe whip restraint, jet impingement barriers, separation, etc.) for safety-related equip-ment at each postulated break location.
b)
Discuss how the main steam lines are protected from tornado generated missiles or how their failure will not damage safety-i related equipmei.:: by failure of the auxiliary or control building roofs on safety related equipment in these buildings.
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410.13 Discuss how safety-related equipment in compartments which contain high or
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(3.6.1)
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moderate energy lines have been protected or qualified to consider the worst case accidental environmental conditions. A crack equivalent to the flow area of a single ended pipe rupture should be used to determine the resulting compartment conditions.
If equipment is to be qualified, provide the temperr.ture, pressure, humidity, potential flooding consequences, and the duration of thes effects t'o'be considered.
410.14 The FSAR indicated that the new fuel storage arrangement assuming optimum (9.1.1) moderation will not exceed a K,ff of 0.95 and that K,ff was conservatively calculated assuming,the new fuel storage area was dry or flooded with unborated water. Verify that a X,ff equal to or 'ess than 0.98 will be maintained with new fuel of the highest anticipated reactivity assuming foam, spray, small' droplets or-mist as a moderator.
410.15 Verify that the new i :1 racks have been designed to withstand the maximum (9.1.1) uplift forces expected for the racks.
410.16 Regarding the safsmic design of the spent fuel storage pool, we nesd the following (9.1.2) information:
a) The FSAR indicates that the pool will be lined with a stainless steel liner plata to ensure water tightness; however, the seismic design of the liner plate is not provided. Discuss why a failure of the liner plate resulting from.an SSE will not result in radioactive release
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from one of the following: mechanical damage to the spent fuel, sig-nificant lott of water from the pool which could uncover the fuel, loss of ability to cool the fuel due to flow blockage caused by portions of the liner plate falling on top of the spent fuel, and damage to safety-related equipment as a result of the pool leakage.
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b) The FSAR does not discuss whether the gate used to separate the spent fuel area from the cask loading area was designed to seismic Category I requirements. The seismic category of the gate should be documented.
If the design does not meet seismic Category I requirements, discuss how a failure of the gate as a result' of an SSE will not result in similar -
conditions as ' staled for the pool liner in part a) of this que: tion.
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410.17 Provide the results of load drop analysis for the primary auxiliary (9.1.4) building crane and the resulting conclusions. This analysis should include the shipping cask and all other loads handled by this crane and an evaluation of the effects of dropping these loads anywhere along the crane's path of travel where unacceptable damage to safety-related equipment could occur.
i Consider equipment on and below the fuel handling bay.
Include the new fuel vault and racks and the spent fuel loading pit.
410. 18 Provide a description of the anti-siphoning methods employed in the design (9.1.3) of the spent fuel pool cooling system to prevent draining of the fuel pool assuming a pipe or component failure.
410. 19 Provide the results of a load drop analysis for the polar crane and the (9.1.4) resulting conclusions. This analysis should include all loads handled by this crane and an evaluation of the effects of dropping of these loads anywhere along the crane,'s path of travel where unacceptable damage to reactor. coolant system components or fuel could occur.
If it cannot be demonstrated that adverse affects to primary system components or fuel will not occur as a result of dropping loads from the containment polar crane, then the crane design must be in accordance with~ the guidelines of Branch Technical Position ASB 9-1 or Regulatory Guide 1.104.
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410.20 Regarding non-seismic Category I components which are served by the Essen-(9.2.1) tial Raw Cooling Water System, provide the isolation signals which are pro-vided to insure isolation of these components under accident condition. The signals are not presently indicated in the FSAR.
410. 21 Your response to question 010.5, describing the use of a non-safety grade (9.2.2) audible alarm to indicate loss of component cooling water to the reactor coolant pumps is not acceptable. As stated in question 010.5, the entire instrumentation system, including audible and visual status indicators for loss of component cooling water should meet the requirements of IEEE Std. 279.
Furthermore, this system and its conformance to IEEE Std. 279 requirement should be addressed inSections 9.2.2 and 7.5 of your FSAR, respectively.
410.22 In order for the staff to perform an assessment of the Ultimate Heat Sink's (9.2.5) ability to meet Regulatory Guide 1.27, provide the results of an analysis of the thirty-day period following a design basis accident in one unit and a normal shutdown and cooldown in the remaining unit.
In suomitting the results of the analysis, include the following information on a day-by-day basis in both tabular and graphical presenta-tions:
a) The total integrated' decay heat.
b)
The heat rejection rate and integrated heat rejected by the station auxiliary systems,-including all operating pumps, ventilation equip-ment, diesels, spent fuel pool makeup, and other heat sources for oath units
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- c) The heat rejection rate and integrated heat rejected due to the sensible heat removed from containment and the primary system.
di The total integrated heat rejected due to the above, e) The maximum allowable inlet water temperature taking into account tne rate at which the heat energy must be removed, cooling water flow rate, and the capabilities of the respective heat exchangers.
f) The required and available NPSH to tr.e essential service water pumps at the minimum Ultimate Heat Sink water level.
Use the methods set forth in Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling," to establish the input due to fission product decay. Assume an initial cooling water temperature based on the most adverse conditions for normal operation.
410.23 Discuss how safety-related equipment will be protected against flooding (9.3.3) assuming a total pipe rupture for all non-seismic piping systems and components (such as tanks) located in safety-related areas or in close proximi ty. Such piping systems should include the fire protection " system.
For the modulating atmospheric dump valves, indicate their failed position.
410.24 (10.3)
Describe the means to operate these valves from the control room considering loss of motive power. If control of the valves is not available from the control room, assuming loss of motive power, comit to verify by test 1
manual cperation of the valves.
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' 410.25 Regarding the bidirectional main steam isolation valves,'inorder for the (10.3) staff to evaluate the design change from steam isolation / check valves to bidirectional isolation valves, provide the following information:
a) With the aid of drawings, describe and discuss the principles of operation for these valves, b) Provide test data which demonstrates the acceptability of this valve under the intended service conditions.
c)
Indicate the nuclear operational experience of this valve as a main steam isolate n valve.
410. 26 Discuss the measures taken at Bellefonte to resolve the potential safety (10.3) concern.regarding the temperature effects on the main steam safety valve setpoint as a result of the main steam safety valve house ventilation system design reported by TVA in NCR SQN MEB 8002 and NCR WBN MEB 8005 for the Sequoyah and Watts Bar plants.
410.27 Section 10,47,, discusses a Heat Rejection System conduit break. Expand (10.4.5) the information provided to include an evaluation regarding the effect of possible circulating water system failure inside the turbine building; Include the following:
a) The maximum flow rat'e through a completely failed expansion ' joint.
b) The potential for and the means provided to detect a failure in tha Heat Rejection System.
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' c) The time required to stop the water flow (time zero being the instant of failure) including all inherent delays such as operator reaction time, drop out times of the control circuitry and coastdown time of pumps.
d) For each postulated failure in the Heat Rejection System give the rate of rise of water in the associated spaces and total height of the water when the water flow has been stopped.
Indicate maximum flow rate for rupture panels in.he exterior west walls of the turbine building.
e) For each flooded space provide a discussion, with the aid of drawings of the protective barriers provided for all essential systems that could become affected as a result of flooding.
Include a discussion of the consideration given to passageways, pipe chases and/or the cable-ways joining the flooded space to the spaces containing safety-related system components.
Discuss the effect of the flood water on all submerged essential electrical systems and components.
410.28 Provide a response to our April 24, 1980 letter (Enclosure 2) concerning your (10.4.9)
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auxiliary feedwater system (AFS) design.
This response should include the following:
a)
A detailed point-by-point review of your AFS design against the Standard Review Plan Section 10.4.9 and Branch Technical Position ASB 10-1.
b) A reliability evaluation similar to that performed for operating plants (refer to Enclosure 1 of the April 24, 1980 letter).
c) A point-by-point review of your AFS design, technical specifications and operating procedures against the generic short term and long term requirements discussed in the April 24, 1980 letter.
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- i d) dn' evaluation of the design basis for the AFS flow requirements and verification that your AFS will meet these requirements (refer to of the April 24, 1980 letter).
410.29 Describe the plant design or operating procedures to preclude secondary (10.3)
(10.4.7) side (feedwater) flow. instability (water hammer) during normal operations
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including startup and shutdown. Also, discuss the potential for water hammer during accident or transient conditions.
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ENCLOSURE 2 p rerg\\
UNITED STATES i
(.7! 4,.,9,E NUCLEAR PEGULATORY COMMISSION i
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'.,g" April 24, 1980 TO ALL PENDING OPERATING LICENSE APPLICANTS OF NUCLEAR STEAM SUPPLY SYSTEMS DESIGNED BY BABC0CK AND WILC0X
SUBJECT:
ACTIONS REQUIRED FROM OPERATING LICENSE APPLICANTS OF NUCLEAR STEAM SUPPLY SYSTEMS DESIGNED BY BABC0CK AND WILC0X RESULTING FP0M THE NRC BULLETINS AND ORDERS TASK FORCE REVIEW REGARDING THE THREE MILE ISLAND UNIT 2 ACCIDENT In our letter of September 27, 1979 to all pending operating license applicants concerning followup actions resulting from our reviews regard-l
-ing the Three Mile Island Unit 2 accident, we indicated that each appli-cant wculd receive additional guidance from the NRR Bulletins and Orders Task Force.
This guidance'would be related to (1) Auxiliary Feedwater (AFW) systems, and (2) analyses for small break loss-of-coolant accidents and in-adequate core cooling, including guidelines for emergency operating procedures.
The purpose of this letter is to advise you of the information we require 4
related to Auxiliary Feedwater s;,ctems.
The requirements were identified during the course of the NRR Bulletins and Orders Task Force review of pressurized water reactor nuclear steam supply systems in light of the Three Mile Island Unit 2 accident.
4 Auxiliary Feedwater (AFW) Systems
.The Three Mile Island Unit 2 accident and subsequent investigations and studies highlighted the importance of the AFW system in the mitigation of transients and accidents.
Following the Three Mile Island 2 accident, operating plants having nuclear steam supply systems designed by Babcock and Wilcox were shutdown.
During these shutdowns short-term actions were taken to improve the reliability of the AFW system. As part of the long-term requ1rements of the shutdown of these plants, more systematic relia-bility analysis of the AFW system has been conducted by the licensees.
The staff is currently evaluating these analyses.
Also as part of its assessment of the Three Mile Island Unit 2 accident and related implications for operating plants, the staff evaluated the reliability of AFW systems for all operating plants having nuclear steam supply systems designed by Westinghouse and Combustion Engineering.
The objectives of the staff's study, related to operation of pressurized water nuclear steam supply systems was (a) to identify necessary changes in AFW system design or related operating proceoures at operating plants in order to as.wa &nda"ad ""~-"---C"7^'-
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Entire document previously entered into system under:
ANO Q C D 9>N N D l
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