ML19319C784
| ML19319C784 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/02/1978 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Stewart W FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003040744 | |
| Download: ML19319C784 (7) | |
Text
.
-r
..,,, I
..,x.
g
/
...x
[
2
..v.
m
..u;~w^
+ -:
.; ' 4; -
- ' a. ;.
A. -m e.
. ~...
c.
_q DISTRIBUTION: /
m- " P:
.e
~
1:.
. Docket ' File t ';~.. Gray File ' -
~
/
x May.2, 19 W J ?!RC ' PDR.
C-
PCheck
~7
~ '
m L PDR LShao
~~
,0RBf4 Rd'g LBarrett.
c
~
Docket No.: 50-302 VStello RBaer BGrimes Attorney, OELD~
OISE (3)
RReid Florida Power Corporation
......CNel son
,1 AlTH: Pr. 9. P. Stewart
....PJogran m....
Director, Power Productjon
,,..DEisenhut..
P. O., Box 14042, Mail Stop C-4
...t...,.JBAbernathy _.
St. Petersburg, Florida. 337N...;.,_MBuchanan..
c ACRS (16):
Gentlemen:
...,2,
.....TCarter-On April 6, 078, we met witEyou.regardini,the recent failure of.
burnable poison cd asserblies.(SPRA).at. Crystal. River. Unit.i!o. 3.
Based on the discussions at this. meeting.we.have.dctcrnined.that_..
the additional infomation requested in the. enclosures should. be.
submitted. identifies.that..information which.should be.
submitted for our review prior to core. reload while; Enclosure E identifies information to be. reviewed prior.to. restart.
It. should
- l be noted that as more detailed.information regarding.this. incident is received, additional questjons y,ay.be necessary.-..
~
c You are requested to submit the information,in. Enclosures -1~.and...
2 as soon as-possible. - Also.vithin_seven. days..please. provide....;.
your schedule for submittal of_the requested.information.as well as for core reload, steam generator repairs,.and. reactor restart.
Sincerely,
~
- NE H : P n
Robert W. Reid, Chief Operating Reactors Branch #4 _...
Division of Operating Reactors.
^
Enclosures:
W
."' 2... ;-.....,;, ;. '
l n Requests for Add?tional,
4,.. i. e, i.'. J;A ~
c r-
' t W-Infbmation gi, X,6.;c..O..... ; '
~J.
s p u:.
v
~
+
z y, g,; c;-v, W.e%,9 ~, w
,9..
,, 4 9.g -
'r s cc w/ enclosures: ' See next pagea. ; ;,..a, Jf;. _. u.. _ _
r t
pm
~+
w-m g,
- ' yp
- k
,ll ly g],, pu/s b r
ORBM: DOR,/C-EB.:)dMI//
C-RSB:00RL Af;Ef3;p0R C-0RBf4.:0 gg.
w%
vv my e wet e,
u o.,
_rgAhn rm JLshan on2.
m L 2 rs-s e+
oomia.
.s T e"^'
.p,
- .5;g;47g.. ;
} su, 7 m y 5yy7g 1
- xi &
- _5jk[73 ; ; s _ 5j.g.j.7g -
,,y n'
.,,4
-u
.u...c.~.w - =.on. w wman m% ncx one D. 0 R* = 4 v.
..====. or==ii
- .m ;%. 1K'
. s f..
A ma 4m ;.*4McCvAw.
.A.wM s GA.4c ;
3 0 4 o M' ".G%.
w M, 4 E
s,
')
Florida Power Corporation cc: Mr. S. A. Brandimore Vice President and General Counsel P. O. Box 14042 St. Petersburg, Florida. 33733 Crystal River Public Library Crystal River, Florida 32629 i
i 1
1
(
a I
l 4
\\
f i
J 1
1
p
- s I
ENCLOSURE 1 INFORMATION REQUEST CRYSTAL RIVER UNIT NO. 3 BURNABLE POISON R00 ASSEMBLY FAILURES PART I Prior to placing the hal back into the reactor the licensee should, as a minimum, consider the following general areas:
1.
Inspection for damage and debris in the primary system.
2.
Recovery of debris and adequacy of cleanup and repair programs.
3.
Effects of any unrecovered debris on reactor safety.
4.
Mechanism that caused BpRA failures.
5.
Design changes required to eliminate future failures at this reactor.
The licensee is requested to provide the folicwing information in sufficient time to allow NRC staff review before placing che fuel back into the reactor.
1.
Describe cleanup operations for removing debris from the primary coolant sistem. This should include a description of any grappling, flushing, /iltration, and vacuum cleanin You should also describe which method (s)g techniques to be used.
will be used for each com-ponent (e.g. fuel assemblies, reactor internals, steam generators, piping, valves,etc.).
2.
Describe the cleanup inspection procedures and techniques which will be used. This description should include any methods used to identify the absence of residual dct. is and the capabilities of the inspection techniques to identify the debris.
3.
Describe the dicage inspection procedures and techniques used.
Identify which components will. be inspected for damage, and what criteria will be used to determine the acceptability of any com-ponents found damaged.
.n
.._._ _._ _ _ _ _.... ~_._
a PART I (CONT'0) 4.
Provide the results of the cleanup and inspections discussed above.
Itemize the total debris recovered and any debris that is not recovered.
5.
Determine the potential effect(s) that residual poison and metallic fragments will have on plant operations. As a minimum address the following areas:
a.
Flow blockage of fuel assemblies.
This should include a con-servative estimate of channel blockage at the end fittings, grids, and in between grids.
You should address the potential for DNB and local cladding hot spots which may cause cladding perforations.
The potential for propagation of fuel failures and the means of monitoring and/or mitigating,such conditions should also be discussed.
b.
The potential for blockage and/or binding of the control rod drive systems due to residual coolant debris. Any procedures planned to mitigate and/or monitor these conditions should be provided, c.
Blockage cf the' guide tubes which nould prevent control rod insertion and safe shutdown operations.
d.
Mechanical damage to primary internals due to impacting.
e.
Blockage and/or binding of any orifices, valve seats, and vent valves in the primary coolant system.
f.
Blockage and/or erosion of steam generator tubes.
g.
The effects that the residual debris will have on pumps and any other components with moving parts.
h.
Effects on coolant chemistry ant :rud levels.
6.
Identify the cause of the BPRA failure addressing possible manu-facturing, design, or installation errors.
Please include:
a.
A description of the "as found" condition of all BPRA in the reactor. Address any indications of improper seating or wear.
i l
t i
.m._,_..
.i.
m._
PART I (CONT'D) b.
Details of nondestructive inspections of the BPRA's, both damaged and undamaged, and orifice rod assemblies.
Address any anomalies found with the holdfown, latch assemblies.
c.
A description of any destructive examinations that have been performed. Address any metalography that has been completed in the areas of wear.
7.
In your presentation on April 6,1978, you indicated that the poison rod assembly was lifted out by action of the hydraulic forces within the core. Provide your analysis of this phenomena.
The complete analysis should include any simplifying assumptions, conservatisms, and tests results used in your evaluation of this phenomena.
Describe what provisions are being considered to preclude this condition and how these provisions will effect other plant operations.
8.
Also during your April 6,1978 presentation, you indicated that the orifice rod assemblies were lifted by the action of the hydraulic forces.
Your basic assumption as to why these assemblies did not experience failure was that they are considerably lighter than the hydraulic forces and therefore are in, essentially, continuous contact with their restraints.
This condition was assumed to eliminate, or minimize, the impact (fatique) damage that resulted in failure of the poison rod assemblies.
If this it true, provide an analysis on the effects that low flow operations will have on the orifice rod assemblies.
9.
Describe the remedies planned to prevent future occurrence of similar failures.
l l
.~
p ENCLOSURE 2 INFORMATION RE0 VEST, CRYSTAL RIVER UNIT NO. 3 BURNABLE POISON R00 ASSEMBLY FAILURES PART II Prior to restart of the reactor the licensee should, as a minimum, address the following areas:
1.
The impact of changes to the core on safety analysis and methods of operation.
2.
Methods to detect future failures.
3.
Any new surveillance deemed necessary as a result of this incident.
4.
Steam generator integrity.
5.
Personnel exposures.
The licensee is requested to provide the following information in sufficient time to allow NRC staff review prior to restart.
1.
Describe how any assynmetries in the core exposure distribution will be accounted for in predictive core physics calculations and in the incore monitoring routine.
2.
Describe how the uncertainty in the exposure of the two fuel assemblies from which the burnable poison clusters were uncoupled will be accounted for in predictive calculations and in incore monitoring.
3.
Provide a safety analysis for operation with the burnable poison removed. This analysis should address the following:
a.
Effects on the moderator temperature coefficient.
b.
Effects on the planar radial peaking factor and on the enthalpy-rise peaking factor.
- c. ' Effects on axial peaking factors, " flyspeck" analyses, and axial. imbalance limits.
Give quantitative information on the axially-dependent depletion of the burnable poison.
~
p PARTII(CONT'0) d.
Effects on shutdown margin and control rod insertion limits.
e.
Effects on core flow distribution.
4.
Describe the testing program to be performed at startup.
5.
Describe quantitatively the methods which will be used to detect further failures (e.g., flux tilts, incore maps, cleanup system, loose part monitor).
In particular, to what impact energy, on the reactor vessel head and steam generator tube sheet and upper head, do the loose-part monitor alarm setpoints correspond?
6.
Describe the damage to steam generator tube sheets and steam generator tubes.
7.
Describe your plans to repair the damaged components of the steam generators including the procedures for rewelding, 1rinding or milling, testing and inspection.
8.
Provide the results of the inspections, tests, and repa'rs discussed above.
'9.
Estimate'the total occupational exposure which will be accumulated in the repair effort.
Your response should include a description of the work to be performed in high radiation areas, measured dose rates in and around the steam generator, time estimates in the radiation fields, and effectiveness of any dose reduction techniques which will be used (e.g., shielding, decontamination, personnel training and special tools) to keep occupational radiation exposures "As Low As Reasonably Achievable." This information should be provided as soon as possible and in time for our review orior to steam generator recair work.
10.
Provide an astinate of all offsite releases which may result from the repair effort.
Such releases may include liquid waste from decontamin-ation solutions or airborne particulates from grinding.
Indicate what treatment systens will be used to reduce levels of radioactivity in plant effluents.
1 l
_,