ML20010E175
| ML20010E175 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/26/1981 |
| From: | Rausch T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8109030171 | |
| Download: ML20010E175 (15) | |
Text
.
Ccmmsnw ith Edison
.y one F ast N eional Phra CNcaga lihnois S
fD }y Address Reply to Post Off,ce Box 767 Chicago, Ilknois 60690 l
' Y! l' August 26, 1981
/
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3 Mr. Harold R.
Denton, Director C
'EP 0 21981
- C Office of Nuclear Reactor Regulation 1
u.s.g%gp g!-
U.S.
Nuclear Regulatory Commission
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Washington, DC 20555 s c,
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Subject:
Quad Cities Station Units 1 and 2 Transmittal of Supplement 3 to Revision 1 of the Licensing Report on High Density Fuel Racks NRC ulcket Nos. 50-254/265 Reference (a):
T.
^
Ippulito letter to J.
S. ADel dated May 15, 1981.
(b):
T.
A.
Ippolito letter to J.
S. Abel dated May 18, 1981.
(c):
T.
A.
Ippolito letter to J.
S.
Abel i
dated May 19, 1981.
(d):
T.
A.
Ippolito letter to J.
S.
Abel dated June 16, 1981.
Dear Mr. Denton:
Enclosed is Supplement 3 to Revision 1 of the report prepared by Joseph Oat Corporation for Commonwealth Edison entitled
" Licensing Report on High Density Spent Fuel Racks for Quad Cities Units 1 and 2".
This supplement provides responses to a portion of l
the questions provided in References (a), (D), (c) and (d).
These responses are for questions numbered as follows:
12.1.5 (Ref, a) 12.2.la, c, d,
and e (Ref. b) 12.2.Sa, b,
c, d,
e, and f (Re f. b) 12.3.4 (Ref. c) 12.3.7 (Ref. c) 12.4.1 (Ref. d) 12.4.2 (Ref. d) l\\p0 0 s t 'M 8109030171 910826 DR ADOCK 05000254 p
H.
R.
Denton August 26, 1981 Please address any questions you may have concerning this matter to this office.
One (1) signed original and thirty-nine (39) copies of this transmittal are provided f or your use.
Very truly yours l'. & &
g
[ ! Thomas J.
Rausch Nuclear Licensing Administrator Boiling Water Reactors Enclosure cc:
Region III Inspector - Quad Cities lm 2458N l
l r
j 4
12.
RESPONSES TO NRC QUESTIONS Given below are NRC questions concerning the Licensing Report on High-Density Spent Fuel Racks for Quad Cities Units 1 and 2.
They are listed by date of 1
transmittal.
Also given below are responses to those questions or the word "Later" indicating that the reasponse will be communicated at a later date.
12.1 Questions from T. A.
Ippolito to J. S. Abel transmitted on j
May 15, 1981 12.1.1 Question:
As a result of replacing the fuel pool racks, there is an appreciable increase in the applied load to the fuel pool concrete floor. Indicate the method and the code used in the analysis of the concrete fuel pool slab.
Response: Later 12.1.2 Question:
Provide the floor response spectra or the time history used in the analysis of the spent fuel racks and state the source of this information.
()
l Response: Later 12.1.3 Question:
)
Iriicate the damping value used in the analysis of spent fuel racks and state whether this value conforms with Regulatory Guide 1.61.
Pesponse: Later i
12.1.4 Question:
Indicate whether material, fabrication, installation, and quality control conform with the ASME code, Subsection NF.
Response Yes, material, fabrication, inspection and quality control conforms with ASME code, Subsection NF.
O 12-1
12.1.5 Question:
Indicate if there is any pocsibility that the shipping cask may drop onto the fuel pool JCner or on to the iuel pool racks and what design considerations are given to the fuel pool liner and racks.
R3sponse: Section 10.1.2 of the Quad-Cities FSAR describes the fuel pool structure and leak detection system.
In regard to cask drop this section references the Dresden-2/3 FSAR (Dockets 50-237/50-249)
Amendment 16/17, Section 11, Fuel Pool Damage Protecticn.
In response to NRC question 2.9.3.11, Section 10 of Amen $wnt 11 of the Quad-Cities FSAR describes the fuel pool liner detien and additional details of the leakage detection system.
Dresden Special Report No. 28 transmitted to the NRC from Commonwealth Edison by letter dated May 31, 1973, provides a structural analysis which concludes that a dropped cask will not penetrate the bottom of the pool. This report also applies to Quad-Cities. Addanda Nos. 1
& 2 trar.smitted to the NRC by letters dated July 2,1973 and August 3
10, 1973 provide additional information.
Modifications have been made to the Reactor Building crane handling system which preclude postulated drops of a 100-ton-spent fuel shipping cask.
These modifications are described in Quad-Cities Special Report No.
16 transmitted by letter from Comnonwealth Edison Company to the NRC dated November 8, 1974.
Supplementary information was transmitted to the NRC by letters dated June 10 and December 8, 1975 and February 9, March 2, and March 29, 1976. The NRC approved the modifications and associated changes in the Technical Specifications in the letter of Januarf 27, 1977 to Commonwealth Edison Company.
12.1.6 Question:
Provide the names of the codes and standards used in the fuel pool liner design.
Response: Later 12.1.7 Question:
With regard to the fuel assembly drop on the top of the rack, provide the following:
a.
Detailed description of the method used to satisfy the accept ance criteria for dropped fuel accident I and II.
b.
Comparison between drops in the tilted position, straight drop and on the corner of the rack.
c.
Indicate whether other modes of failure of the racks exist beside crushing.
Response: Later 12-2
12.1.8 Question:
Indicate in detail the methodology used to demonstrate the leaktight integrity of the fuel pool liner when subjected to either the postulated fuel assembly drop or the cask drop over the spent fuel pool liner. The heavy drop should be analyzed for the tilted position and straight drop.
Response: Later 12.1.9 Question:
Indicate whether the proposed fuel storage pool modifications conform with the staf f position on " Fuel Pool Storage and Handling Application", dated April, 1978, including revisions dated January, 1979.
If any deviations exist, identify and justify these deviations.
Response: Yes, the guidance is followed, with the exception of the Technical Specification for maximum enrichment.
This is because of the variety of enrichments in the fuel and the existance of the subcriticality specification of k less than or equal to 0.95.
eff 12.1.10 Question:
The seismic analysis as presented in the submittal is not lear. Indicate in detail how all the seismic models and parameters (Figure 6.1, 6.3, 6.4, 6.5, 6.6, 6.7 and 6.8, the friction forces and floor response spectra) fit together to predict the seis rte stresses.
Indicate the interrelationship among the models.
Response: See Revision 1 to Chapter 6,
Seismic Analyses Description, submitted to the NRC by letter fron T. J. Rausch to H. R. Denton on June 24, 1981.
12.1.11 Question:
Because different type modules were used in the proposed modification with different sizes and weights, indicate which type was used in the seismic and sliding analysis.
Indicate also how other types were qualified for the postulated loadings.
Response: Later t
n i
12-3 1
=
_ _ _ _ = _
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12.2 Questions from T.
A.
Ippolito to J. S. Abel transmitted on May 18, 1981 12.2.1 Question:
i When Section 5.1, Heat Generation Calculations, is provided, include the following information:
I a.
Indicate the minimum elapsed time between shutdown and when the dis charged fuel is in the spent pool for all anticipated fuel discharge cycles.
Reponse: See Section 5 of Supplement 2 to Revision 1 of the Licensing Reports submitted to the NRC by letter from T. J. Rausch to H. R.
Denton 3
dated August 10, 1981.
b.
For Units 1 and 2 spent fuel pools, indicate the number of fuel assemb lies and their respective decay times of all fuel that will be in the pools when reracking occurs.
Response
See Revision 1 of Licensing Report submitted to the NRC by letter i
from T.
J.
Rausch to H. R. Denton on June 24, 1981.
c.
It is noted in the FSAR that portions of the RHR system may be used to augment the spent fuel pool cooling system by inserting spool pieces in the spent fuel pool cooling lines shown in Figure 10.2.1.
In this regard, indicate the length of time required to install these spool pieces and describe the capability of the RHR s,' stem to remove the heat from the spent fuel poo' over a range of pool temperatures and with and without the spent fuel pool cooling system in operation.
Response: The time required to install the spool pieces is discussed in the response to question 12.2.2.
The capability of the RHR system to remove heat from the spent fuel pools is discussed in Section 5 of 3
Supplement 2 to Revision 1 of the licensing report, submitted to the NRC by letter from T.
J.
Rausch to H.
R.
Denton dated August 10, l
1981.
d.
For Units 1 and 2 indicate the length, width and depth of the spent fuel j
pools and the minimum volume of water in each when all storage racks are l
filled with fuel assemblies.
{
Response As shown in Fection 2 of the licensing report, the length and width I
of each pool are 41 feet and 33 feet respectively.
The depth of water in each pool is 39 feet. As stated in Section 5 of Supplement 2 to Revision 1 of the licensing report, submitted to the NRC by 3
letter from T. J. F.ausch to H. R. Denton dated August 10, 1981, the water inventories in the Quad-Cities Unit 1 and Unit 2 spent fuel pools are 44887 and 44471 cubic feet respectively when all racks are in place in the pools and every storage location is occupied.
O 12-4
_ ~ _.., - ~. _. _
. _ _. _ _ _ _ =.
1 1
1 e.
Figure 2.1 and 2.2 of the March 26, 1981 submittal shows that the down-comer region, i.e.,
space between the racks and walls of the pool, is quite small. Further, the vertical dimension of the water plenum formed t
by the base plate of storage racks and the pool bottom is 6-1/2 inches.
Assuming the maximum heat load is adversely located in the storage racks demonstrate that sufficient circulation will occur to preclude nucleate boiling.
Response See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, 3
submitted to the NRC by letter from T.
J.
Rausch to H.
R.
Denton l
dated August 10, 1981.
12.2.2 Question:
1 Assuming the reactor is operating at power when it becomes necessary to utilize the RHR system to cool the spent fuel poo'., describe and disc.uss the steps that must be taken and the elapsed time before the RHR system can be placed in the fuel pool cooling mode of operation.
Responss Using the Residual Heat Removal (RHR) System for fuel pool cooling will render one of the two loops (two pumps and one heat exchanger) unavailable for use in any of the safety functions (LPCI or containment cooling).
Quad Cities Technical Specifications allow LPCI and one loop of containment cooling to be inoperable during reactor operation as long as 1) the other loop of containment cocling is available, both core spray systems are operable, and both diesel generstors are operable, and 2) the loop used for fuei pool cooling is returned to normal within seven days, or the reactor shall be shut down.
Once it has been determined that supplemental fuel pool cooling using RHR is necessary, the RHR/LPCI Mode Outage Report i
Surveillance would be performed, and crews would be dispatched to install the two spool pieces which join the fuel pool cooling system to RHR. When this has been accomplished, the valving cperations may begin. This involves the closing of several motor-operated valves, racking out the breaker on another motor-operated valve, and the opening of two manual valves near the fuel pool cooling heat l
exchanger s.
Next, the RHR Shutdown Cooling Mode suction header must be filled end vented and the RHR systert vented. Finally, the RHR service water system is started and an RHR pump is started to commence fuel pool cooling.
The total elapsed time would be approximately three hours if two maintenance crews were available (cne for each spool piece) or four hours if a single crew installed both spool pieces. At times when no riaintenance crew is on site, an I
additional one to two huurs would be required to assemble the necessary personnel.
O 12-5
l l
i 12.2.3 Question:
1 For both Units 1 and 2 spent fue pool reracking operations, provide the j
following additional information:
a.
Assuming a load drop, describe and discuss, with the aid of drawings, the i
travel paths of the new and existing storage racks with respect to plant equipment that may be needed to attain a cold safe shutdown or to d
j mitigate the consequences of an accident, l
Response: Diagrams will be prepared before moving racks based upon results of NUREG-0612 studies.
b.
Provide the weights of the racks. Describe and demonstrate the ade quacy of the lifting rig attachment points, on the new and old racks, to i
withstand the maximum forces that will be eroerienced during the load handling operations.
Response: The weight of the racks is contained in the Revision 1 Licensing Report submitted to the NRC on June 24, 1981 by letter from T. J.
Rausch to H.
R. Denton.
Lifting rig reqttirements at e not yet defined and will be submitted later.
I c.
With the aid of a drawing, describe th e lifting rigs that will be employed in handling the racks and demonstrate their adequacy.
Response: Later i
d.
Assuming stored spent fuel is in the pool when the storage racks are being removed or installed, demonstrate that the stored spent fuel is net within the area of influence of dropped racks should one or more of legs of the lifting rig fails.
Response: Later i
e.
FSAR Figures 12.1.1 and 12.1.2 shows a transfer canal joining Unit 1 pool with Unit 2 pool. Assuming a significant number of loads are transferred between the two pools, describe the merits of providing additional protection in the form of a cover over those storage racas directly under this frequently travelled path.
l Response: The assumption that a significant number of loads will be trans-ferred between the pools is incorrect. Both pools are nearly full which precludes significant transfers of fuel.
With regard to adding a cover, this cover would only add another heavy object consideration in additon to thermal cooling concerns, i
I l
O 12-6 l
()
f.
For both Units 1 and 2, with the aid o' drawings, sequentially describe the movement of the stored spent fuel assemblies and storage racks in order to reduce the possiblity of fuel damage in the event of a load drop during the reracking operations.
Response All work will be planned in advance and detailed procedures de-veloped to reduce the possibility of load drops and resaltant fuel damage.
g.
Considering the limited space between the storage racks and the pool walls, describe the travel paths and laydown area for various pool gates.
Demonstrate that the consequences of a dropped gate are acceptable or that one can reasonably assume that dropping of the gates is very unlikely.
Response: Later h.
Using Figure 3.7, describe and discuss tbs ability of the high density storage racks to protect the stored spent fuel assemblies from damage following a load drop.
Response: Later i.
In regard to the potential for damage to stored spent fuel resulting from light load drops (i.e., one fuel assembly and its associated handling tool when dropped from its maximum carrying height), it was assumed that
()
all lesser loads that are handled above stored spent fuel would cause less damage if dropped. Verify that this assumption was correct, e.g.,
indicate that all lesser loads when dropped from their maximum elevation would impart less kinetic energy upon impact with the tops of the fuel assemblies arid or storage racks.
Response: Later j
12.2.4 Question:
l Since Figure 2.2 shows that essentially all available space in Unit 2 pool will be occupied by storage racks, therefore, all Unit 2 stored spent fuel must be moved to Unit 1 pool via the transfer canal before it can be loaded into tne shielded shipping cask. Describe and discuss what measures will be taken to reduce the possibility of fuel assembly damage resulting from the additional fuel handling operations.
Response: It will not be necessary to move cli Unit 2 fuel thru the Unit 1 pool whea it becomes possible to ship fuel. The rocks in the Unit 2 l
cask handling area will not be installed unless required.
If they j
were installed, they could be removed to facilitate the use of a cask later. In addition, all fuel movements will be accomplished by approved procedures to reduce the possibility of tuel assembly damaga.
12-7 L
12.2.5 Question:
For both Unit 1 and Unit 2 storage pools, starting with the total decay heat load that will exist in each pool following the reracking operations, provide the following information:
a plot of the pool's maximum anticipated total decay heat load result ing a.
from normal discharges versus time until each pool has reached its storage capacity.
Response: Decay heat loads for several limiting cases are discussed in Section 5 of Supplement 2 to Revision 1 of the Licensing Repor t, 3
submitted to the NRC by letter f rom T. J.
Rausch to H.
R.
Denton dated August 10, 1981.
b.
Verify that all decay heat calculations have been made in accordance with ASB technical position 9-2.
Response: All decay heat calculations have been made in accordance with 3
Branch Technical Position APCSB 9-2 (now ASB 9-2).
c.
a plot of the pool's water temperature versus time for each discharge where the total decay heat exceeds the capacity of the spent fuel pool cooling system. Indicate what cooling systems are in operation and their respective capacities.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, J
submitted to the NRC by letter from T.
J.
Rausch to H.
R.
Denton dated August 10, 1981.
d.
a plot of maximum decay heat load in each pool, assuming a full core discharge at each of the normally scheduled refueling periods.
Response: See Section 5 of Sepplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter f r om T.
J. Rausch to H.
R.
Denton dated August 10, 1981.
d e.
a plot of the pool's water temperature versus tine following each full core discharge assumed in Item d above.
Indicate what cooling systems are in operation and their respective capacities.
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter from T.
J.
Rausch to H.
R.
Denton 3
dated August 10, 1981.
O 12-8
i 2
1
! ()
f.
Assuming the maximum heat load exists in Unit I and Unit 2 pools wban all external cooling was lost, indicate the time interval before boiling
]
occurs and the boil off rate.
l l
Response: See Section 5 of Supplement 2 to Revision 1 of the Licensing Report, submitted to the NRC by letter from T.
J.
Rausch to R.
R.
Denton 3
{
dated August 10, 1981.
I g.
Desc ibe and discuss the sources of makeup water, the quantity avail 1
able, their respective makeup rates and the steps that must be carried j
out and the elapsed time before the makeup water will be available at the pools.
j Response: Later 4
12.2.6 Question:
Since the RHR system will be required to augment the sps.nt fuel cooling systed for some period of time following a discharge, describe and discuss how it will be verified that the decay heat load has decayed to a value within the capacity of the spent fuel pool cooling system and, therefore, allowing the j
RHR system to be safely returned to its safety function mode of operation.
l Response: It has been CECO's experience that the RHR is not required for i
either a reload or full core discharge.
It was required, its use would be phased out by throttling back the RHR and observing if the O
pool temperature remains stable. If it is stable, the spool pieces would be removed and the RHR returned to its safety function.
i l
12.3 Questions from T. A.
Ippolito to J. S. Abel transmitted on May 19, 1981 1
12.3.1 Question:
I i
Discuss in some detail, the procedure that will be used for (1) removal of the l
fuel rods from the present racks, (2) removal and disposal of the racks j
themselves (i.e.,
rating them intact or cutting and drumming them), (3) l installation of the new high density racks and (4) loading them with the l
presently stored spent fuel rods.
In this discussion include, in a step by l
step fashion, the number of people involved in each step of the procedure l
including divers if necessary, the dose rate they will be exposed to, the time l
spent in this radiatica field and the estimated man-rem required for each step of the operation.
l Response: Later i
O 12-9 l
l
()
i 12.3.2 Question:
Demonstrate that the method used for removal and disposal of the old racks will providt ALARA exposure.
Response: Later l
12.3.3 Question:
l What r3diation levels will be used to determine whether the racks to be disposed are identified as clean or radioactive racks.
Response: 1000 DPM per em is considered clean.
12.3.4 Question:
Identify the important radionuclides and tQggr (4 ci/cc) in the spent fuel pool water including Cs,qggentCs, 5g neen ragf ns i
p Co, and Co.
What is the external dose equivalent (DE) rate (mrem /hr) from these radionuclides. Consider these DE ratos at the edge and center of the pool.
]
Response: See Section 8 of Supplenent 2 to Revision 1 of the Licensing Report, submitted to the NMC by letrCr from T.
J.
Rausch to H.
R.
Denton 3
dated August lf., 1981.
12.3.5 Question:
O Provide an estimate of the increase in annual man-rem from more frequent changing of the demineralizer resin and filter cartridge.
4 Response: Ac discussed in Section 8 of Revision 1 of the Licensing Report, the proposed modification will have a negligible annual effect on the pool cleanup system; therefore, there is expected to be no increase in the annual frequency of changing of the filter demineralizer resin.
12.3.6 Question:
oiscuss the build-up of crud (e.g.,
Co, 60Co) along with the sides of the 58 pool and the removal methods that will be used to reduce radiation levels at the edge of the pool to ALARA.
Response: A buildup of crud as a result of this proposed modification would mean that the concentration of crud in the pool water has increased.
Because the cleanup system removes essentially all crud deposited in the pool water from one refueling long before the next refueling, a measurable buildup will not occur.
(See Section 8 of Revision 1 of the licensing submittal.)
In addition, operating experience to date indicates no significant buildup of crud along the sides of the pool.
O 12-10 1
h 12.3.7 Question:
Provide ari estimate of the total man-r 7m to be received by personnel oc-cupying the spent fuel pool area based on al' operations in that area in cluding those resulting frem 4, 5,
and 6 above.
Describe the impact of the modification on these estimates.
Response: As discussed in revised Section 8 in Supplement 2 of Revision 1 of the Licensing Report, there is expected to be negligible to no increase in man-rem as a result of the modification.
Assuming a radiation dose of 4 mr/hr around and above the pool (see Section 8 3
Of Sapplement 2 to Revision 1 of the Licensing Repor t) and occupancy of 50C0 man-hour daring refueling and 4000 man-hour /yr at other
- tires, the total exposures are 20 an-re=
and 16 man-rem /yr respectively.
12.3.5 Nest;on:
!3entify the c ni t :r i ng syste s that will te c'e 3, s.s3 its 1ccation in the spe-t f;al pool
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12.4 Questions from T. A.
Ippolito to J.
S. Abel transmitted on 1
June 16, 1981 1
12.4.1 Question:
Describe the samples and instrument readings and the frequency of measurement that are performed to monitor the water purity and need for spent fuel pool i
cleanup system demineralizer resin and filter replacement. How will these be l
affected by the proposed e-tion?
Response: Water purity is monitored by a continuous conductivity meter installed on the inlet to the fuel puol demineralizers, and by periodic grab samples for laboratory analysis.
Once a week a representative grab sample is obtained from the fuel pool demineralizer inlet line.
The analyses performed are pH, chloride, silica, and turbidity.
The activity checks are gross beta and gross alpha counts.
3 Once a month a sample from the same location is obtained for a gamma isotopic analysis.
All major peaks are identified.
All identifiable isotopes are cuantified, and an LLD is determined for Kr-85.
The criteria for a demineralizer backwash and precoat is a consistent excursion frcm the chemistry
- limits, or high differential pressure across the demineralizer. Each demineralizer has differential pressure instrumentation installed which will alarm in the Unit's control room and the radwaste control room if a preset value is exceeded.
The proposed change is not expected to alter the chemistry or radiochemistry of the spent fuel pool; consequently, the described measurements will not be changed.
12.4.2 Question:
l State the chemical and radiochemical limits to be used in monitoring the spent i
fuel pool water and initiating correcting action.
Provide the basis for establishing these limits, giving consideration to conductivity, gross gamma and iodine activity, demineralizer and/or filter differential pressure, demineralizer decontamination factors, pH, and crud level.
1 i
Response: The chemical and radiochemistry limits used in monitoring the spent fuel pool water are as follows:
Conductivity
< l.0 # mho/cm pH 6.0 - 7.5 3
< 0.500 ppm Silica
< l. 0 ppm Turbidity None Gross Beta
< lE-02 F Ci/ml Gross Alpha
< lE-05 # Ci/ml i
12-12 1
I, s
j If any of the above limits are exceeded the recomrranded action is to i
backwash and precoat the fuel pool demineralizer.
1 1
The basis for the water chemistry limits is the G.E. Water Quality document (22Al286, Rev. 0) that provides the water specifications for various plant systems. The limits are set to minimize corrosion and to maintain the water in a "cryrtal clear" conditien, t
3 i
The radiochemistry limits have been established based on operating
{
experience as action levels below which personnel exposure in the vicinity of the spent fuel pools is minimized.
The demineralizers are backwashed if differential pres-are exceeds j
25 psid for protection of the filter elements.
(
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I I
1O 12-13 i
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a