ML20211G941
| ML20211G941 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 08/23/1999 |
| From: | Murray B GENERAL ELECTRIC CO. |
| To: | Mendonca M NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9909010070 | |
| Download: ML20211G941 (12) | |
Text
.
GENuclear Energy i
, , , GenewDemc Compam j
YdIIHCitOS huCleJ? Certer
- LAugust 23,1999 sus v#ces seas sanct cA Sssas t .
.Marvin M. Mendonca, Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate '
{
Division of Reactor Program Management l Office of Nuclear Reactor Regulation
- U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 !
References:
- 1. Docket Number 50-73, License R-33,
- 2. " Request for Additional Information (TAC No. MA0099)", Marvin M. Mendonca to G. L. Stimmell, dated October 20,1998.
- 3. NEDO-32740," General Electric Nuclear Test Reactor Safety Analysis Report",
August,1997.
- 4. Letter, B. M. Murray to Marvin M. Mendonca, dated June 18,1999.
Dear Marvin:
The responses to the remaining questions of the request for additional information (Reference 2) {
regarding the General Electric Nuclear Test Reactor Safety Analysis Report (NTR SAR, Reference 3) is
{
submitted in the attachment to this letter. The reply to each question follows an italicized restatement of l the question / comment. The partial response was submitted on June 18,1999 (Reference 4).
The responses included in this submittal address the following page numbers: 2-11,4-1,4-25,5-4,5-9 &
10,7-15, Il-2,12-17 (Note: An expanded response to the question regarding page 12-17 is included.),
13-8,13-9, and 13-37.
The following NTR facility figures from Chapter 10, in response to the comment on page 4-12, are enclosed: Figures 10-1,10-2,10 3,10-4,10 5.
If there are additional questions related to this response, please contact me at (925) 862-4455.
Very truly yours, bk '
B. M. Murray Senior Licensing Engineer Attachments cc: Mr. Steve lisu Radiologic Health Branch State Department oflicalth Service P.O. Box 942732 t'}
70 I Sacramento CA 94234-7320.
n 1 A 'T o o P
m ,
. R Pane 2-11: As describe!! earlier in the Safety Analysis Report, the site is surrounded by barren
' mountains and rolling hills. Provide an analyticaldescription ofthe effects ofthis terrain on the effluent releasesfrom the stack in the meteorology section or other appropriate section.
To our knowledge, there has been no requirement for, nor has there been an analysis of the e ffects of the regional terrain on the releases from the VNC effluent stacks. Environmental sampling on and around the VN_C site has been conducted for over 30 years. The reported results of the sampling program provide assurance that there has been no observed reconcentration of radioactive efiluents from the VNC operations.
The stack effluent release limits are conservatively derived from, among other parameters, the annual average dilution-dispersion coefficients. The dilution-dispersion coefficients were calculated from meteorological observations from a two-year period. Ground level release points at the stack locations were used as one of the criteria in the calculations. The calculated release limits for a particular stack were further reduced by a factor of two to account for releases from other stacks on the VNC site. The use of a ground level release provides conservatism over the use of an elevated release in the calculated release limits. All stack release points at VNC are located above ground level.
Measured and reported releases from VNC stacks are consistently less than 10% of the annual average release limits, where the release limits are established to comply with the 10CFR20 effluent concentration limits. In fact, the particulate releases are typically below detection limits of the VNC counting laboratory, or they consist of natural Radonffhoron progeny. Continued use of environmental gamma dosimeters and air sample stations, in conjunction with the continuous stack monitoring system, verify compliance with site and public dose limit requirements. In the highly unlikely event of a design basis release, the emergency plan calls for special monitoring, including air sampling in the down-wind direction from the release. This would be the basis for additional actionstif needed.
Revisions have been made to the SAR, Section 6. Replace pages 6-8 and 6-9 with the enclosed pages.
Paec 4-1: Since the current core " container was put into service in 1976 after the previous i container, which had been in servicefor approximately 18 years, sprung a leak in a weld area," provide an analysis of the current condition of the container. Include potential weld and materialneutron embrittlement considerations. Also, according to the discussion on page 4-10 the configuration allows inspection ofthefuel container. Provide the results ofthese inspections. In addition, provide reasoningfor a TS requirement onfrequencyforfuture
- inspections.
In this regard, please discuss,from an aging perspective, the current condition of other safety related reactor components such as control and safety rods and their drive mechanisms, wiring, coolant piping and other safety related wiring and relays, including relay contacts. If inspections of these components have been performed, please provide the results.
I 1
1 The reactor core can has been replaced twice: once in 1976, and the other time shortly after NTR first went critical. Both failures were in weld areas and not due to corrosion. While it is possible to perform an inspection of the core can, it would be an extremely time consuming and high radiation exposurejob. It would require production of specialized equipment that does not currently exist. Because of ALARA considerations, regular inspection is not scheduled nor performed. The consequences of a leak in the core can are minimal. Leakage is detected by the presence of water on the floor. The reactor cell sump would collect any leakage and cause an alarm in the security building (occupied 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day). A float switch in the fuel loading tank will also cause an alarm in the security building upon a low-level condition due to a leak. If all water were to leak from the core can, no fuel damage or dangerous radiation conditions would exist, even with the reactor operating. An alarm from either the sump high level or fuel loading ;
tank low level would be cause for reactor shutdown or scam. Loss of primary flow or high !
temperature core outlet water will initiate an automatic scram signal. Reference section 13.4.6 for the analysis for a loss of primary water accident. j The effects of aging on other reactor components are discussed in Section 16 of the SAR, Components are inspected and tested periodically and appropriate maintenance is performed.
For instance, safety rod magnets and springs have been replaced. Safety rod and control rod I wiring is inspected and replaced when needed. Control rod drive belts have been replaced when j slippage is noted. The NTR is a simple machine; it operates at a low temperature and pressure, l
and the consequences of anticipated operational occurrences and postulated accidents, such as a l core can leak, are minimal. These events are operational problems rather than safety problems.
4 Pane 4-25: Providefurther detail on the derivation of thepeakingfactors. Provide a description or referencefor what combination ofexperimental and analytical methods to derive these peakingfactors. Similarly, provide an explanation of"with neutronflux peaked on one side of the core. " Describe the measurements and/or analyses used to establish circumferential and axialpower profiles. Compare the results of these derivations to the assumptions used in this safety analysis report.
Radial and axial peaking factors were measured values obtained from the files. The term
" neutron flux peaked on one side of the core" is an awkward way of saying radial peaking.
Peaking factors used in the analysis, then, include the following:
radial 1.25 axial 1.15 other 1.1 The "other" peaking factor includes all other factors such as may be caused by fuel uranium-aluminum alloy non-homogeneous effects, fuel dimensional tolerances, and meter reading tolerances.
The total overall power peaking of 1.58 is the conservative simultaneous occurrence of all these effects and is obtained by multiplying the three factors together, i.e.,1.25 x 1.15 x 1.1 = 1.58.
2
l 1
I Paec 5-4: Regarding the allowable prirnary systern leakage rate of10 gallons / day, provide or
- reference, a calculation of the maximum radiological exposure to radiological workers and to the public considering the maximum allowable radioactivity in the primary coolant system.
l Also, a similar calculation should be madefor the potentialloss ofcoolant accident analysis in ,
Chapter 13. The radioactive source term should be consistent with the statement on page 5-9 i that "(a)n inadvertent release ofexcess radioactivity in theprimary coolant, ofhigh enough l
level, would cause the reactor cell remote area monitor to alarm. " That is, the level of \
radioactivity in the primary shmdd be thatjust below what would initiate this alarm, or at a level corresponding to other indications that would terminatefacility operation and initiate corrective actions.
A simplistic evaluation can be made to estimate the activity that would have to accumulate in the l' reactor cell (assumed to concentrate in a point source) that would cause the remote monitor to alarm (at 100 mrlh). Using the following assumptions, the activity of mixed fission products ;
released from the primary system causing a 100 mrlh alarm can be estimated using the "R=6CE" rule of thumb (where R is the dose rate at one foot, C is the activity in curies, and E is the average energy of the radioactive material):
Assumptions: i Distance from the source to the detector = 30 feet. j l
Average energy of the mixed fission products = 0.7 MeV ;
R, at one foot = (30/l)2 x 0.1 R/h = 90 Rlh l Then, l
The activity necessary for a 100 mrlh alarm = 90/(6 x 0.7) = 21.4 curies. l
. The radiological effect of this quantity of released fission products, or other radioactive material,
)
from the primary water would be negligible to the public and environment. There is no normal !
condition mechanism for release of this activity from the reactor containment building. The l effluent from the stack exhaust must pass through a HEPA filter before exiting the stack. The !
stack monitoring system would detect abnormally high stack particulate (or gaseous) releases. !
Leaks from the primary system would normally migrate to a sump, from where the accumulated l water would automatically be pumped to a wastewater tank. Disposal of the contaminated l wastewater would be controlled using either a deionization system or the site evaporator. l Normal monitoring and handling procedures would maintain worker exposures AL. ARA and l within regulatory limits.
To further quantify the magnitude of the activity required to cause a high radiation alarm, the following volume estimation is made:
The measured gross beta-gamma activity in the primary coolant ranges from approximately 0.03 pCilml at shutdown conditions to 1 Ci/ml during reactor operation. !
The volume of water necessary to deposit 21.4 curies of gross beta-gamma activity from a core can leak would range from approximately 188,000 gallons to 5,700 gallons, respectively. This volume of water far exceeds the 1,830-gallon capacity of the primary '
water system (core can, piping and fuel loading tank).
3
Therefore, the maximum quantity of gross beta-gamma activity available to accumulate on the i
cell floor is approximately 7 curies. The hazard associated with this quantity of radioactive material is localized to operators entering the cell, and essentially nil for public exposure.
Note: Short half-life nuclides present in reactor water during operation are not considered in this evaluation.
Paec 5-9 & 10: Provide or reference the analyses that demonstrate that the specyication on primary coolant water conductivity ensures aluminum corrosion is within acceptable levels.
The reference for pH and corrosion rates for aluminum in water is:
C.R. Tipton, " Reactor Handbook, Volume 1, Materials", Interscience Publishers, New ,
York (1960), p. 861. l Corrosion occurs at a slow rate over a period of time. Relaxing the conductivity limit, then, for brief periods allows time to correct a condition causing the high conductivity while not contributing to accelerated corrosion. The magnitude of the allowed conductivity increase and the brief time period specified would not cause increased corrosion of the system.
1 Paec 7-15: Provide a description ofthe leak testingprocedures and requirementsfor the j neutron source. Include consideration offrequency and whether or not a Technical i Specification requirement is needed.
Excessive personnel exposure would be required to perform direct smear tests of the radium-beryllium neutron source. The sealed source is located in an aluminum access tube in the graphite moderator. The dose rate near the entrance to this tube is typically greater than 1,000 mrem /h. Retraction of the source would require time in this dose rate field, and the dose rate would increase further if the source were removed from the tube. Performance of a direct smear test is contrary to an ALARA practice. The source has no direct contact with reactor coolant or other water. Thus, there is no water-induced corrosion mechanism.
Articic 30275(d) of the State of California Code of Regulations, Title 17 allows test sampling of the holder, or from the surface of the device in which the source is stored or mounted, or an alternate test for contamination and leakage may be approved by the Department of Health Services. Routine smears of the area surrounding the access tube are taken and analyzed by the Regulatory Compliance personnel and the VNC counting room. This routine smear test is considered to meet the intent of a sealed source leak test, while minimizing personnel exposure in compliance with ALARA.
Pane 11-2: Provide a description ofor reference (e.g., section 16.1) to the other methodsfor detectingfuelleakage other than theprimary coolant sampling. Provide an analysis to demonstrate the sensitivities ofthese other methods to detectfuelfailure. This analysis should demonstrate that the combination of these other methods and the annualprimary coolant sampling are acceptable to detect and mitigate operation with leakingfuel to prevent unacceptable radiological exposure orfuel damage.
4
I Other than primary coolant sampling, fuel failures may be detected by the on-line primary water d
. conductivity meter, the radiation level on the primary coolant demineralizers, increases in the reactor cell area radiation monitor or radiation levels outside the reactor cell, stack effluent monitoring, and observations of reactor power oscillations or power drop.
Since the reactor is open to atmosphere, radioactive gasses released to the primary system will diffuse to the reactor cell and, consequently, to the stack. Released particulate material would i affect the conductivity and be trapped in the demineralizers. The radiation levels on the demineralizers are generally maintained at less than 2 R/hr and are routinely recorded and monitored. At higher radiation levels, the demineralizer resins are scheduled for replacement.
There are no anticipated operational occurrences or postulated accidents that could result in fuel damage.
Pane 12-17: Provide verification in thefirstparagraph ofsection 12.9 that the QA program l includes managerial and administrative controis to ensure safe operation in accordance as required. That is, QA is not limited to design and construction. Also, iduntify the differences between this QA program and the guidance in ANSUANS 15.8 " Quality Assurance Programs for Research Reactors." Provide an analysis toJustify the differences or adopt the guidance of ANSUANS 15.8.
The QA Program provided in Section 12.9 of the SAR applies to design, purchase and i installation of hardware for the physical plant. QA at NTR does include administrative and I managerial controls to assure safe and reliable performance by the operators performing operation and maintenance of the facility. These requirements are contained in other documents as allowed by Section 3 of ANSI /ANS-15.8. Specifically, each Facility Operations requirement I of this Standard and the NTR requirement are listed below: !
- Organization ANS 3.1 is contained in NTR Technical Specification 6.0.
l
. QA Program ANS 3.2 is contained in Vallecitos Safety Standards, NTR SOPS, Job Descriptions and Performance Appraisals.
- Performance Monitoring ANS 3.3 is contained in Nuclear Safety Procedures. l
. Operator Experience ANS 3.4 is contained in the Requalification Program and implementing SOPS.
. Operating Conditions ANS 3.5 is contained in NTR operating and maintenance procedures.
. Operational Authority ANS 3.6 is contained in NTR SOPS as applicable for a single shift operation.
. Control Area ANS 3.7 is contained in NTR SOPS.
- Ancillary Duties ANS 3.8 is contained in NTR SOPS.
i j e Emergency Communication ANS 3.9 is contained in the Site Emergency Manual and l in NTR SOPS.
- Configuration Control ANS 3.10 is contained NTR SOPS.
5
- Lockouts and Tagouts ANS 3.11 is contained in the Site Facilities Maintenance Procedures.
- Test and Inspection ANS 3.12 is contained in NTR SOPS.
e Operating Procedures ANS 3.13 is included in the NTR SOPS.
- Operator Aid Postings ANS 3.14 is not used.
. Equipment Labeling ANS 3.15 is incorporated as appropriate for the small size and small staffing of the NTR facility.
Pace 13-8: Provide clartfication on the net reactivity characteristics that are said to be shown in Figure 13-2.
The SAR incorrectly states that the net reactivity is shown in Figure 13-2. In fact, the net reactivity is not shown on the graph. The amount of positive reactivity that could be added in this anticipated operational occurrence is less than 0.10$. This amount of positive reactivity is not significant and the inclusion on the graph would not contribute to understanding the dynamics of the event. This variable is not included.
l Paec 13-9: Figure 13-2 seems to assurne a punty startfrom 100 k W at time zero that is not f allowed by Tech Specs. Provide additional explanation of Figure 13-2 to describe the transient or conditions it is depicting.
Figure 13-2 does show this anticipated operational occurrence with an initial power level of 100 kW. Reactor operation under these conditions is not allowed by Technical Specifications, and the safety system would additionally cause a reactor scram under these conditions. The scram would be caused by the low flow scram or in a few seconds by the high temperature scram.
IIowever, for conservatism, the initial conditions were chosen to be full power with shutdown temperature conditions. These initial conditions conservatively result in a higher peak temperature and a larger positive excess reactivity addition due to temperature change. The analysis shows that even for these incredible conditions, the consequence of the event is benign.
Pnec 13-37: Provide stack draft characteristics tofurther demonstrate the conservatism in the concentration calculation as indicated in thefootnote.
The height of the release point was one of the criteria used in the calculation of boundary doses that could result from exposure to radioisotopes released from a failed, fueled experiment. The evaluation used a ground release height [ reference item (9) of 13.6.3, page 13-39]. The effect of any stack draft characteristics on the boundary radionuclide concentration calculation was not a factor in the determination of the NTR experiment limits.
6
O
~
~
_I,o' fi
- o.E .
k, *
- O*,. -
o*_l
. ll
~
Y Y - T ._
T IE I E ..
O ER .
ER H DC -
'o DC .
2 HN
- H N G
}
v n GO I C I C ,
H
/ a v
5 " .-
H) -
u s $. .
i ' , , , f GkE : I .. ,-
T , ' -
n tl;.#, 3* m -
Y H
'O.
'S
. ./
~
n i e NPV a
e OAT RReL d ._ utg n i a -
TGs o
'i i e UOC saA
, R R sc i o
R .
LS TL a
_ W C ECE
_-:y% v e
- mU@ ' ?
_ Q - U1 AC E
- e
& "O M . - N R _ l l E
'o bO o
% A 's ~ R R" I - g .
6 t O v h l: gl
,I 4
R H 3. .
i
[ rT NN o"
c d
n a
- T -
n -
% ;. . .'. . ., i i..I,.*r,'r.,A,;:.J'.* j',,
d
- , R ..
l
-
- O . , ' ,
.# , p', i a ll:
)
j N
g l
P
~.
- N 3( d , ,
r 7- -
3 A f o -
E f o Q: '
R l A '/ F G
W -
3 y
K R f
- t r
1 .a
'[ ~ - (U 0O O
W .
P a -
"*22. , ,
!1 *- ,
- u u
23 ;
. L A
f- 1 -
4 u
- R E ,"
0
}. u ,5, / N 1 -
?, o ' - " E u ,
/tg' G -
i.
- l e -
- - - ,. . . *,~, .
j d
r -
'. .'o ,i,.'. , :
- 4. .h'* . c. '.
,g T.,h. s,
.r u
- 's
- l4 .. ' # :
'e ' ".. g f - -
_". i -
r
, F _
. A .
E n R ,
A "4 's ..
,U
- - E Y "2 G , l l
33 - A R ,
O .
T .
S c -, -
_ l ~
U *l
~ ; _-
l .e; 3L
. 'o o~ ,
o.* aol ,6 *. -
I' -
- = -
,0 '-
l 1
l t
REACTOR CELL
, , j a'. O [* h. , . .
- 8 ' '. *:f 1
. y, _. . ;.t , , '
g .i y' " <,; *' ~ ' ' '
a : ~, .,.,.5,a. '
}l_w . ka 4 v 'l.t * , e ;
. e - lj*.' 4:,',' ~, q? ' * '
SHIELDEO CASK i
- p. ***'y;m' .
RECESS
\
g7 7
. l s ,9 g., s l g.7,4 v* . i
- e. ; ... 4 j r fs,,,
I, i .
.ev,.. o'
&"j-}'l . I o;f'/4j
- i~!.-;,"$*,,'jf'
- 2 9
w 4 " HIGH DENSITY
- ~ ML b / SHIELDED CONCRETE
~
(
/
/
(I
\ BLOCKS
/
/n i
IMAGEFOll i
\
TRANSFER MECHANISM /
tinoirect neutron radiography) %
N FILM CASSETTE ANO LEAD SHIELO TRANSFER MECHANISM (direct neutron radiogroony - explovve location) NORTH ROOM Figure 10-2 Modular Stone Monument Neutron Radiography Facility
o .
NCRTH ROOM *
- DENETR A TIONS {
MggR "o,., PCR C&l FACILITY ANQ I
MONUMENT, % 8 INSTR UMENTATION s' LEACS 7.f* y;'.e
',, .55 '! ..-, e
'16(.
.*!J!;.
%;Yii.; ' ' ' ' '.: .,.
.D,. . :iT ~
,.,y Vp.: l,{ ' , ,.f, .:.
l- :._,jg,.l:
. . .'M
.;. saUrTER j' of::. $yy, .e^ .. s l MCLCUP TANK .
W$i i CONoutTS TO CONTROk ROOM e
o : );, "' ' l l
REACTOR CONTROL C hA
, ELEMENis N o o j$.5: ? Jg.7 R EMc V A86E o o ..
.N. ..i REACTOR CELL' o 0c
- N o
/ (PLVG ELEV.10 f t)
MAIN .M. .
O oRAnwirE ,,
y g g y,gg g PACK ' ,. . p . ,
{
~ . /,,c,y ry FUtL t REMOVASLE
' Losonno TAnM N
- ** , . '. + " **..e
==="* pgyo
..* wear p CORE em .e.4 ,
is, ele
.. 7 ERMAL cooas To,;ay, *
. / CoLuuN gg:
~
c d.wwwad7; a . e - *. .*
- g... ** . 4..' ENERoy i
d.. f..?gt.t.*ik e.
. . . ..**. **0'
?, ARsoRFrioN 8 "' 5 'S 1
",ifE #@jlhI8y ze . *
.****I- - aoR8ZoNTAL l ' :2 . / a s P*g*:t o +
M t i.af e, lgf*,*N; 5*r*
f.***.= ? 4, . , *:
1 FAC:u tv
- ooCooo ** fr i
- y CCNQUnTS / :,yv *. 4
- 5 MUTTER y'If
- f. ..'.
i o
s" 7 / reCELL c.
- m. !
,W ?j i J.w'. ,*
l ,
i, .* e.g l $
! h5 T*.IR NUCLEA R .- SCUTH ! II CONSOLE *
'gp. CELL CCNTROL E
N?.th_
.?*
?h Roou * -
,,4c 3 rr a-' ? -
- g y?;
Eas? WALL
- e , , ,, *EnttRafton
.e . v.
L *~
. . 5?sY.t.u *I '*" '*
o .5*.Iw.. *..$. N.!*a'. .- E?t
. C$n'"CR ern - . . . . . . - --
G C5b" LIE Y YE utao
'p,% nsRArvnn GRAPmft Figure 10-3 : uclear Test ReaC:or Tacilities (TCp View)
l.
{
EXPERIMENT TUBE
( __
_}
) HORIZONTAL 4" x 4" \
( / PENETPATIONS )
bha $ N hY4h ,
M /*
\ /08 pfW*
( -E -g
< i 3 ) 1 "
5" HORIZONT A IL TY s
f U 2"x2" EQUIPMENT 7 PENETRATION
) AREriusawas=o REMOVAELE )
e -
l Figure 10-4 Horizontal Penetrations
. 6 e
a e
- _
- g y g
4 C
a g
g yu -
o r
c x
, A M
gg g g gg g
gcg gyA
. gg , p ie t y r g - s y a ,
- . t ct t N A*t o
S k
es
)
uAl s g l
l t l a A;
- t s
i N
t8 K M _
ND
- OA
- Ni t _
g-g m_*-_;- 3 I' , _
_- g
_ a g
g gg y
c q % g 3
c g
A ya g _
g - _
- e _
g i i a _
g g lp c a g
y _- r t
Cl o
t A
t
_- l d
N a
_- t I
84 _- t M
58 1
8
_- L I
A ne r
N S
_ _- u t
8 5 N i ts .s
/ -
i E
s a
- H r
e s
/ - _
t c
_N / L a
_ _ t y7 4
g g
y n1 f
5,
.s.9
- e. .
_ _ e m f da g
, i c
. . s.- ~
_ c ,
y s
a g N Tg
_ i
- Cgg e * ,g
[t elb% *- -
o!
j g a g
tc J
t A
- g IN y 0 5
.V.*-
t gygg g 1
_. g S _
s, 3a yit .
w gA l A;
cct 3 -s t
AA cit -s I _
s B t H
D:
Aet m AG
) l mi f 3 D s
i t
-W t i 4
4 _ i t N S AA u ! t 4 F 3 - _ S I 4
4 3 .t y - _
l il H
D Y'
9 -
- ,.e s
N -
r 6
9 Q :.
.4
)
M s
@ ' :e n4 g.,yo,,- :n:3; t
- i