ML17338A874
| ML17338A874 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 06/11/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| References | |
| NUDOCS 7907240546 | |
| Download: ML17338A874 (20) | |
Text
I JUNE 1 1 1979 Docket Nos. 5-an
-25'l Dr. Robert E-Uhrig, Yfce President Advanced Systems and Tec{tnology Florida Power and Light Company Post Office Box 529100 Miami, Florfda 33'152 Dear Dr.
Uhrfg:'UBJECT:
IHFORNATIOH COPY OF RE(UEST FOR. ADDITIONAL INFORI'$ATION REGARDING WESTINGHOUSE SMALL BREAK LOCA ANALYSIS As we discussed with you during.our meeting an Hay 30, 1979, the NRC staff is evaluating certain aspects of the NRC licensing process in light of the Three Nile Island Unit 2 (THI-2) accident of f5arch 28, 1979.
One of the aspects being addressed by t{ie staff in its evaluation is the status of the models used in safety analyses, especially with regard to transients and small break LOCAs.
Enclosed is a copy of an NRC letter, dated June 4, 1979, requesting information fram your nuclear steam supply vendor, the {lestfnghouse Electric Corporatfan, regarding small break LOCA analyses and analysis methods.
This enclosure is provided for your information.
It is our understanding that you are participating in this effort as it applies to your licensed power plants through an o;iners group.
At some point in this effort >>e will require that you document the, appIfcabflity of the re'suits of these Mestfnghause analyses to your plants.
To that end, a copy of all correspondence to Mestfnghause and the oNners group on this matter will be provided.to you and placed in aur files on your speciffc plants.
Please contact P.
D. O'Reilly, the staff's assigned project manager for this matter, if you have any questions.
Qrigina{ Signed By gp,75.84I ~~
A. Schwencer, Chief Operating Reactors Branch 81 Division af Operating Reactors E
0IlIllC C+1f OVRNAMCW closure:
G June 4;" I97 to {/estfnghou oaTc~
I...,N ~AKIM
~ Iet ter-'"":".. DOR.: ORB1........
f{Grotenhufs'Oct.."
l.79---"
DOR'ORBt bASahv(eneer.
~ 96"" 795-"-
BRC PORM 318 IL'age (976) NRCM 0240
+ 4,5 OOVCRNMCNT 5NINTINO Ol ~ ICCI I 555 COILS 5 ~ 5
'h
'I h
QC" t L
Fg I" 1
~ I II[
E v
v
~y% AE0(
%p
.0 Docket Nos.
50-'250 and 50-251 UNITEDSTATES NUCLEAR REGULATORY COMMISSION V4ASHINGTDI4,D. C. 20555 JUNE 1 1 'l979 Dr. Robert E, Uhrig, Yice President Advanced Systems and Technology Florida Power and Light Company Post Office Box 529100 Miami, Florida 33152
Dear Dr. Uhrig:
SUBJECT:
INFORMATION COPY OF REQUEST FOR ADDITIONAL INFORMATION REGARDING WESTINGHOUSE SMALL BREAK LOCA ANALYSIS As we discussed with you during our meeting on Hay 30, 1979, the NRC.
staff is evaluating certain aspects of the NRC licensing process in
'ight of the Three Mile Island Unit 2 (TMI-2) accident of Harch 28, 1979.
One of the aspects being addressed by the staff in its evaluation is the status of the models used in safety analyses, especially with regard to transients and small break LOCAs.
Enclosed is a copy of an
.NRC letter, dated June 4, 1979, requesting information frcm your nuclear steam supply vendor, the Westinghouse Electric Corporation, regarding
.small break LOCA analyses and analysis methods.
This enclosure is provided for your information.
't is our understanding that you are participating in this effort as it applies to your licensed power plants through an owners group.
At some point in this effort we will require that you document the applicability of the results of these Westinghouse analyses to your'lants.
To that end, a copy of all correspondence to Westinghouse and the owners group on this matter will be provided to you and placed in our files on your specific plants.
Please contact P.
D. 0,'Reilly, the staff.',s assigned project manager for this matter, if you have any questions.
Sincerely, 1
Encl osure:
NRC June 4, 1979 letter to Westinghouse cc:
w/enclosures See next page A." Schwencer, Chief Operating Reactors Branch "1
Division of Operating Reactors
f
Robert E. Uhrig Florida Power and Light Company JUNF 1 1 lb>o cc:
Mr. Robert Lowenstein, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N.M.
Suite 1214
~
Mashington, 0.
C.
20036 Environmental and Urban Affairs Library Florida International University Miami, Florida 33199 Mr. Normyn A. Coll, Esquire
- Steel, Hector and Oavis 1400 Southeast First National Bank Building Miami, Florida 33131 Mr. Henry Yaeger, Plant Manager Turkey Point Plant Florida Power and.Light Company P.
0.
Box 013100 Miami, Florida 33101 Mr. Jack Shreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304
gp,S REGS
~4 O
e An I
': ~
C IO "I
+>>**+
UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JUN
~
~~79 Mr. Tom M. Anderson, Manager Nuclear Safety Department Westinghouse Electric Corporation P. 0.
Box 355 Pittsburgh, Pennsylvania 15230
Dear-Mr. Anderson:
SUBJECT:
RE/JEST FOR ADDITIONAL INFORMATION REGARDING SMALL BREAK LOCA ANALYSIS As you are aware, the NRC staff is evaluating certain aspects of the NRC licensing process in light of the Three Mile Island Unit 2 (TMI-2) accident, of March 28, 1979.
One of the aspects being addressed in this evaluation is the status of models used in safety analyses, especially those used to analyze transients and small break LOCA's.
In order, for the staff to complete its evaluation of the response of currently operating Westinghouse-designed plants to postulated small break LOCA's, additional information is required.
These information needs, which pertain to expected system behavior following postulated small break LOCA's and small break LOCA analysis methods and results, are identified in the enclosure to this letter.
Please provide, within seven.
days of the date of'eceipt of this letter, your schedule for responding to the items contained in the enclosure.
We recognize that some of the requested information may have already been provided to the staff in the course of other reviews.
In lieu of refer-encing already submitted material, we would prefer, in order to facilitate our review, that the responses be provided in one complete document.
The information contained in the enclosure was discussed with your repre-sentatives at our May 31, 1979 meeting.
If you have any question regarding the contents of this letter, please contact P.
D. O'Reilly, the assigned project manager.
Sincerely, D.
F. Ross, Jr.,
Deputy Director Division of Project Management Office of Nuclear Reactor Regulation
Enclosure:
As stated
ENCLOSURE RE EST FOR ADDITIONAL INFORMATION REGARDING SMALL BREAK LOCA ANALYSIS The response of the primary system of a given plant to small break LOCA's will differ greatly depending upon the break size. the location of the break; mode of operation of the reactor coolant
- pumps, numbers of ECCS systems functioning, and the availability of secondary side cooling.
In addition, this response may differ for different plants designed by the same NSSS vendor because of differences in loop configuration (two-loop, three-loop, or four-loop) or different ECCS designs.
In order for the staff to complete its evaluation of the response of currently operating Westinghouse PWR designs to postulated small break LOCA's, the following information is needed:
{1)
Provide a qualitative description of expected system behavior for (a) r a range of postulated small break LOCA's, including the zero br'eak case, and {b) feedwater-related limiting transients combined with a'tuck-open power operated relief valve.
These cases should include situations where auxiliary feedwater is both assumed available and not available.
The cases considered should also include breaks large enough to (a) depressurize the primary system, (b) maintain the primary system at some intermediate
- pressure, and (c) repressurize the primary system to the safety and/or relief valve setpoint pressure Yarious break locations in the primary system should be considered, including the pressurizer,
{2)
Provide a qualitative descriplion of the various natural circulation modes of expected system behavior following a small break LOCA.
Discuss a>>y ways in which natural circulation can be interrupted.
In particular, discuss the applicability of the concerns in the Nichelson reports (reports I
on BKW 205 FA plants and CE System 80 plants) identified in 'Annex 1 to this Enclosure.
Assess the possible effects of non-condensable gases contained in the primary system,
The following questions pertain to your small break LOCA analysis methods:
(3)
Demonstrate that your current small break LOCA analysis methods are appropriate for application to each of the cases identified in items (8) through (l2 )below, This demonstration should include an assessment of h
the adequacy of the pressurizer and steam generator
- noding, and the pressurizer surge l ine representation.
This may be accomplished by verifying the methods, with the use of data (e.g.,
comparison with experiments, THI-2 evaluation).
If, as a result of the above assessment, you modify your analysis methods (e.g., pressurizer and steam generator noding), provide justification for any such modification.
/
{4)
Verify the br eak flow model used for each break flow location analyzed in the response to Item '{8) below.
(5)
Verify the analytical model used to calculate natural circulation heat removal under two-phase flow conditions.
(6)
Provide justification for your treatment of non-condensible gases following discharge of the safety injection tanks.
(7)
Verify your analytical calculation of fluid level in the reactor pressure vessel for small break LOCA's and feedwater transients.
S
~ For each of the analyses requested fn Items (8) through (12) below, (i)
Provide plots of the output parameters specified in Annex 2 to this Enclosure.
(ii)
Indicate when the pressurizer safety and/or relief valves would open.
{iii) Include appropriate information about the role of control systems in the course of the transient.
Describe how the system response would be affected by control systems.
(iv) If the scenario is different for different classes of plants (two-loop, three-loop, four-loop, different ECCS designs),
provide an example of each kind.
8)
Provide the results of a sample analysis of each type of small break behavior discussed in the response to item (1) (e.g., depressurization, r
pressure hangup, repressurization).
(9)
Provide the results of an analysis of the worst break size and location in 'terms of core uncovering.
This may be a break which does not result in HPI initiation.
This may require more than one calculation.
(10)
Provide the results of a complete analysis of feedwater-related limiting transients combined with a stuck-open power operated relief valve.
These cases should include situations where auxiliary feedwater is both assumed available and not available.
{11)
Provide the results of a small break LOCA analysis assuming loss of feedwater and auxiliary feedwater.
The case with the worst break location which affords the least amount of time for operator action should be analyzed.
Single failure of the ECCS should be considered.
(12)
Provide the results of a small break LOCA analysis assuming that one steam generator is lost either due to isolation or due to loss of auxiliary feedwater.
(13)
Provide the results of an analysis of. the effect of reactor coolant pump operation (tripping all RCP's, keeping all and some RCP's running) on the course of small break l.OCA's.
('l4)
Provide the results of an analysis of the effects of different HPI termination criteria on the course of small LOCA's.
Specifically, for each small break LOCA ana'}yzed in response to Item (8L %boise.,'cqmpare the effects of the NRC HPI termination criteria (as stated in I8E Bulletins79-06A and 79-06A, Rev. 1, item 7(b)) to those for the HPI termination criteria which you have recommended to licensees with. Westinghouse'esigned operating plants.
Provide plots of significant parameters of interest, I
such as system pressures, temperatures, and subcooling, on a
common time axis.
Indicate on the plot when the operator would terminate HPI. injection for both sets of criteria.
I (15)
Provide a list of transients expected to lift the PORVs; identify the assumed steam and two-phase flow rates through the valves for these transients.
Provide justification for your assumptions,, including the time at which two-phase flow discharge would be experienced, (16)
Provide guidelines for the preparation of operational procedures for the recovery of plants following small LOCA's.
This should include both short-term and long-term situations and follow through to a stable condition.
The guidelines should include recognition of the event, precautions,
- actions, and prohibited actions.
If RC pump operation is assumed under two-phase conditions, a justification of pump operability should be provided.
Discuss instrumentation available to the operator and any instrumentation that might not be relied upon during these events (e.g., pressurizer level).
What would be the effect of this instrumentation on automatic protection actions7
NNEX 1
TVA C. Michelson Concerns C
Pressurizer level is an incorrect measure of primary coolant inventory.
The isolation of small breaks (e.g.,
letdown line; PORV) not addressed or analyzed.
Pressure boundary damage due to loadings from a) bubble collapse in subcooled liquid and 2) injection of ECC water in steam-filled pipes.
In determining need for steam generators to remo.e decay heat, consider that break flow enthalpy is not core exit enthalpy.
Are sources of auxiliary feedwater adequate in the event of a delay in cooldown subsequent to a small LOCA?
Is the recirculation mode of operation of the HPSI pumps at high pressure ah established design requirement?
Are the HPSI pumps and RHR pumps run simultaneously?
Do they share cottmon piping?/suction?
If so, is the system properly designed to accomiodate
>his mode of operation (i.e., are any NPSH requirements violated, etc...?)
ci, tiechanical effects of slug flow on steam generator tubes needs to be
- ddressed.
(transitioning from solid natura'1 circulation to reflux boiling and back to solid natural 'circulation may cause slug flow in the hot leg pipes).
9.
Is there minimum flow protection for the HPSI pumps during the i'ecirculating mode of operation?
- 1.
The effect of the accumulators dumping during small break LOCAs is not taken.
into account.
11.
What is the impact of continued running of the RC pumps during a small LOCA?
'i2.
During a small break LOCA in which offsite power is lost, the possibility and impact of pump seal damage and leakage has not been evaluated or analyzed.
13.
During transitioning from solid natural circulation back again, the vessel level will be unkn'own to the procedures and operator training may be inadequate.
addressed and evaluated.
to reflux boiling and operators, and emergency This needs to be "OTE:
Items 1 through 4 are taken from "Decay Heat Removal During A Very Small Break LOCA for a BKW 205-Fuel Assembly P'WR," C. Hichelson, Draft Report, January 1978.
Items 5 through 15 are taken from "Decay Heat Removal Problem Associated with Recovery from a Very Small Break LOCA for CE System 80 PWR,"
.".. Nichelson, Draft Report, Hay 1977.
(continued next page)
ANNEX 1 page 2
14.
The effect of non-condensible gas accumulation in the steam generators and its possible disruption of decay heat removal by natural circulation needs to be addressed.
15.
Delayed cooldown following a small break LOCA could raise the containment pressure and activate the containment. spray system.
impact and consequences need addressing.
ANNEX 2
Plotted Out ut Parameters C(re:
L, X, T Reactor Vessel:
Lter Head:
L, X
'3owncotter:
L, X
f'ig~n HotLeg:
X, T,M, L (Pressurizer Leg)
Cold Leg:
X, T, W, L, WWP>, jWWP>dt (Break Leg)
P-c=.:urizer:
Win, Xin, L, X, P, T Steama Generator:
Primary:
X, L, T, h
Secondary:
P, L, X, T, WREL, WAFW, h Leak:
'"'iRV, W, X
or
- Break, W, X, Wdt 1'~l.
1 1: X.l Nomenclative:
P - Pressure L-Hixture Level X - Quality T - Temperature W - Mass Flow Rate h - film heat transfer coefficient HPI - High Pressure
!njection REL - Relief Valve AFW - Auxiliary Feedwater.