ML20050P028
| ML20050P028 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 04/12/1982 |
| From: | Hodges M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20050N853 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8204140457 | |
| Download: ML20050P028 (10) | |
Text
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
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In the Matter of
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LONG ISLAND LIGitTING COMPANY
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Docket Nos. 50-322
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l-A NRC STAFF TESTIMONY OF MARVIN W.
.(' AYNE) H0DGES CONCERNING WATER HAMMER d
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OUTLINE OF TESTIMONY Suffolk County Contention 4 asserts that Applicant has not adequately demonstrated that safety-related piping at Shoreham will be able to withst'and the adverse effects of water hammer since it has not considered the start-up experience at similar BWR plants.
Staff witness Hodges disagrees. He testifies that the problem of water hammer has been given generic consideration in various Staff studies. This information has enabled Staff to focus on certain areas of concern. Applicant has provided Staff information regarding these areas and has been denonstrated that steps have been taken to insure that safety-related piping at the Shoreham nuclear facility will not be substantially impaired by water hammer.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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LONG ISLAND LIGHTING COMPANY
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Docket Nos. 50-322
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(Shoreham Nuclear Power Station, Unit 1)
NRC STAFF TESTIMONY OF MARVIN W. (WAYNE) H0DGES ON SUFFOLK COUNTY CONTENTION 4 Q.
Please state your name and position with the NRC.
A.
My name is Marvin W. (Wayne) Hodges.
I am employed by the U.S. Nuclear Regulatory Commission as a Section Leader in the Division of Systems Integration. A copy of my professional qualifications is attached.
(Exhibit A).
Q.
What is the purpose of your testimony?
A.
The purpose of this testimony is to respond to Suffolk County Contention 4 which states that:
Suffolk County contends that LILC0 has not demonstrated adequate assurance of the operability of safety-related piping to prevent or withstand the effects of water hammer because the Company has not considered the start-up experience at similar BWR plants. Therefore, Shoreham safety-related piping (e.g., ECCS, Reactor Decay Heat Removal Systems) does not meet 10 C.F.R. 50, Appendix A, GDC 1, 31, and 46.
Q.
Can the Staff respond for LILC0 concerning the extent of reviews made by LILC0 of startup experience at similar BWR plants with regard to water hamer?
2 A.
The Staff cannot respand for the Applicant since the response would require details of the history of the Applicant's design and review process which are not available to the Staff.
However, the Staff has considered water hanmer in nuclear power plants on a generic basis and has questioned the Applicant concerning water hammer in the Emergency Core Cooling System (ECCS).
Q.
Describe briefly the Staff generic considerations of water hammer.
A.
The generic consideration by the Staff of water hammer in nuclear power plants was incorporated into Task Action Plan (TAP) A-1,
" Water Hammer." TAP A-1 is described in NUREG-0371, " Approved Task Actions for Category A Generic Activities." Work on this plan has included water hanner analyses, state-of-the-art reviews, and a detailed review of water hammer events that have occurred both prior to and af ter 4
commercial operation for both boiling water and pressurized water reactors. The water hammer event review originally covered events up through 1978, but has been updated to the present time.
It should be noted that the Staff definition of water hammer for the event reviews was generalized in meaning to include transients involving steam (steam hammer) and two-phase flow (e.g., water entrainment in steam lines, steam bubble collapse) in addition to the classical water hammer transients such as those involving valve closure and pump startup in solid water systems.
Q.
What information did the Staff obtain concerning water hammer in the ECCS at Shoreham?
. A.
The Staff requested and received information concerning the ECCS design and operation provisions to prevent and mitigate water hammer.
Specific questions were directed to water hammer due to water entrainment in the steam lines of the High Pressure Coolant Injection System (HPCI) and the Reactor Core Isolation Cooling System (RCIC) and to water hammer in the pump discharge lines due to inadvertent voiding.
The Staff reviews had indicated that'these were the two most common causes of water hammer in the ECCS for BWRs. The Staff question and Applicant response are given in Exhibit B.
The response indicates that the Applicant was aware of these two key causes of water hammer and had provided system design features to minimize occurrence of these types of water hammer.
Q.
What are your conclusions regarding this contention?
A.
Contrary to Suffolk County's contention, as has been demon-strated in its answers to questions in Exhibit B an adequate assurance has been demonstrated that the Applicant has taken into account the experiences of other facilities and has provided adequate precautions to prevent water hammer.
6
tsma Marvin W. (Wayne) Hodaes
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Professional Qualifications Reactor Systems Branch Division of Systems Integration U. S. Nuclear Reculatory Commission
- I am employed as a Section t.eader in Section B of the Reactor Systems Branch, DSI.
I graduated from Auburn University with a Mechanical Engineering Degree in 1965.
I received a Master of Science degree in Mechanical Engineering from Auburn University in 1967.
In my present work assignment at the NRC, I supervise the work of 6 graduate engineers; my section is responsible for the review of primary and safety systems for BWRs.
I have ser,ved as principal reviewer in the area of boiling water reactor systems.
I have also participated in the review of analytical models use in the licensing evaluations of boiling water reactors and I have the technical review responsibility for nany of the r.:odifications and analyses being implemented on boiling water reactors post the Three Mile Island, Unit-2 accident.
As a uamber of the Bulletin and Orders Task Force which was formed af ter the 1MI-2 accident, I was responsible for the review of the capability of BWR systcms to cope with loss of feedwater transient and small break loss-of-coolant accidents.
l I have also served at the NRC as a reviewer in the Analysis Branch of the NRC in the area of thermal-hydrulic performance of the reactor core.
I served l
j as a consultant to.the RES representative to the program management group for the EWR Blowdown / Emergency Core Cooling Program.
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Prior to.ioining the !'RC staff in March,1974, I was employed by E. I. DuPont at the Savannah River Laboratory as a research engineer.
At SRL, I conducted hydraulic and heat transfer testing to support operation of the reactors at the Savanr.ah River Plant.
1 also performed safety limit calculations and participated in the develophent of ar.alytical models for use in transient
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ar.alyses at Savannah River.
My tenure at SRL was from June 1967 to March 1974.
frc,m September 1965 to June 1967, while in graduate school, I taught courses in therm 3 dynamics, statics, mechanical engineering msasurements, conputer prograrring and assisted in a course in the history of engineering.
During the su.v.:.er of 1966, I worked at the Savannah River Laboratory doing hydraulic testing.
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EXHIBIT-B
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Request 212.59 (5.5, 6.3) :
Recent event reports f rom operating bWRs have 'shown that water ba,rner may occur in the ECCS during testmg.
There is concern that undetecrea damage caused by this water barr,ner may result in f ailure of an emergency sysmin when needed to perform under accident conditions.
Provide assurance that your ECCS designs and operation incorporate provisions to
- prevent, mitigate, and detect the ef f ects or water ha rroer.
Specific information should also 1;e providea to address the following potential problem areas:
(1) How will water hans.er in the HPCI system be minimized during HeCI startup with a cold steam supply line following a
cold shutdown?
Consider that th e HPCI steam supply isolation valves are closed until reactor pressure is increased above the low pressure isolation setpoint.
This information should be provideci f or RCIC also.
(2) Tae aesign and oI> erat $on of the ECCS fill systems should be suf ticient to keep the discharge lines full or water and inf orm the operator of off-normal conditions.
Provide values for allowable fill system upper and lower pressure lunits and describe tne bas,es used to determine these limits.
Describe the ala r~m i.rovided to infonc the operator or off-normal conditions approaching these li rni ts.
1,ow pressure limit Q
suitches should have sufficient setpoint values and D
sensitivity to detect a
partially full discharge
- line, including the conon_lon involving fill system flow into the ECCS discharge line.
Response
Water hairmer is prevented in the ECCS systems ny ensuring that wah-r lines are normally illled at all times and that HPC1/ACIC turbine steam supply and exhaust lines are contamuously drained.
In those cases where it is impractical to maintain water lines
- dilled, such as the RER head and containment spray lines and the hak t.est return lines downstream of the innermost containment isola tion
- valves, the erfects of water ha:umer are mitigated by periocuing a time history analysis or water hanner loads and aesigning t.h e pipe support system to acccrrmooate these loads.
any postulated erfects resulting from water hancuer during i,CCS testing would be actected during inservice inspection of piping, components,. and supports.
Specific information regarding HPCI.
and-nCIC start-up following a cold shutdown and the design and, operat_ ton of the i.CCS loop -level systems f ollo,ws:
1.
Water hazme.r will be prevented in the HPCI and RCIC systems durin9 start-up by sequential _ly opening the respective systems' steam supply isolation valves.
The HPCI and RCIC steam supply 1hes re.nain isolated until the reactor pressure reacnes 100 and 50 psig, respectively.
At the 100 and 50 psig point.s, the respective HPCI and RCIC valves are 212-59 Revision 11 - June 1978
1 SNPS-1 FSAR opened as follows:
.EPCI
- outboard bypass valve F080 and
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inboard bypass valve F097 are opened to drain any moisture,
y, upstream of valve F002, to equalize the pressur.e across valve F002, and to warm up a line down to the steam supply valve by the turbine.
This warms the steam line up to the temperature corresponding to a 100 psig reactor pressure.
The line is sloped downward to a
drain pot system which removes any condensate which collects when the system is not operating.
Then the inboard steam isolation valve F002 is opened.
- Thus, the line is drained dry and maintained that way at all times preventing water slug damage.
Additionally, is the reactor pressure increases the temperature of the line will increase matching the corresponding saturation points.
-A comparable scenario applies to RCIC..
The outboard bypass valve is F075; the inboard bypass valve is F085.
The outboard and inboard isolation valves are F008 and
- F007, respectively.
Both systems' turbine exhaust lines are sloped to either the suppression pool or the exhaust line drain pot to remove any exhaust line moisture.
2.
RER and Core Spray Systems - Two loop level pumps, shown on the core spray system PSID (Fig. 7.3.1-4), supply line fill water to the discharge lines of both the core spray and the RHR systems.
The two pumps are obviously separated: one pump supplies fill water for core spray loop A and RER loop A while the second pump supplies fill water to core spray
.f{A, loop B and RER loop B.
The fill pumps are backed up by the (2/
condensate transfer system for periods of fill pump maintenance.
These loop level pumps keep the discharge piping between the RHR and' core spray pump discharge check valves and the containment isolation valves, as well as the RHR shutdown cooling line, filled at all times.
RHR and core spray suction piping and discharge piping up to the pump discharge check valves are below minimum suppression pool-water level and are therefore always full following initial system vent and fill.
With the exception of RER reactor pressure vessel head line, containment spray lines, and the RHR test return lines, all piping inboard of the containment isolation valves is maintained in a
full condition from either initial. fill or previous operation.
Loop level pump discharge pressure prevents drainage of these lines back through containment isolation valves.
Pressure switches exist in the discharge of each loop level pump to alarm before the pressure drops below,the point where draining could occur.
The alarm setpoint is 58 1 - 1 psig, l
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'l corresponding to a head of at least 10 ft at the highest
-q) point of potential draining,
which is at the core spray testable check valves.
In addition, flow sw'tches exist in i
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the loop level discharge lines to alarm at high flow (5.0 1 0. 3 gpm).
This setpoint is high enough to allow reasonable check valve leakage back to the suppression pool and will warn the operator of abnormally high drain rates well in advance of the low pressure alarm.. A wide range of adjustment exists in both the low pressure alarms
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to 212-59a Revision 12 - July 1978
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SNPS-1 PSAR o
P 100 psig) and the high flow alarms (0 to 15 gpm) to allow for M.
flexibility in operation and in instrument location.
The loop level pumps discussed above serve no other function; therefore, any abnormal. flow or pressure condition indica'es t
a problem in the core spray or RER system discharge line or in the loop level pump.
In order to ensure that the lines are actually full.and that the pumps are not merely compressing air
- pockets, selected high point vents are periodically opened to verify that the pipes are full.
This high point vent surveillance is performed in accordance with the Technical Specifications.
- These systems each have their own loop level pump to keep the piping filled in a
fashion similar to that descrioed above for the RHR and core spray systems.
The till pumps are backed up by the condensate transfer system for periods of fill pump maintenance.
The level pump switches are set at 40 1 2 psig with a range of 0
to 200 psig.
This setpoint corresponds to a head of at least 10 ft at the highest _ point of potential draining, which is at the feedwater testable check valves.
Flow switches also exist on the loop level pump discharge lines, with the same
- function, setpoints, and tolerances as those in the RliR and core spray syst m.
also incorporate high point vents with the same surveillance described above.
_g The ECCS loop level systems are therefore equippeci with redundant ~ and diverse instru:nentation of adequate range and accuracy to ensure that all required pipes are full.
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212-59b Revision 11 - June 1978
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