ML20002C971
| ML20002C971 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/27/1964 |
| From: | Bryan R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20002C970 | List: |
| References | |
| NUDOCS 8101150505 | |
| Download: ML20002C971 (29) | |
Text
-
1
. f y
t HAZARDS ANALYSIS BY THE RESEARCH AND POWER REACTOR SAFETY BRANCH i
DIVISION OF LICENSING AND REGULATION CONSUMERS POWER COMPANY OF MICHIGAN DOCKET NO. 50-155 i
OPERATING LICENSE REVIEW AND PHASE II 0F THE RESEARCH AND DEVELOPMENT PROGRAM
' It.troduction During tha term of its 18-month Provisional Operating License Consumers Power Company has conducted a power operation test program and has engaged in Phase I cf a Commission-sponsored Research and Development progr'am. Power Operations havc thus. f ar been limited to 157 Mw(t). Consumers has now requested, by 1
-Anendment No.14 to its application dated November 15, 1963, (1) conversion of the provisional operating license to a 40 year license, (2) authority to operate l
the Big Rock Point reactor at its design rating of 240 Mw(t), and.(3) authority to conduct Phase II of the Research and Development Program.
This Hazards Analysis summarizes the staff's evaluation of Consumers application.
In 'particular we have considered the following aspects of the Big Rock Point operations'as they relate to the requests enumerated above:
),
1.
Power operation test program.
2.
Plant operating history.
3.
Results of Phase I, Research and Development Program.
4.
Proposed escalation to 240 Hw(t).
5.
Research and Development Program - Phase II.
Power Operation Test Program i
A special report on the plant operation from initial criticality through full power demonstration, dated August 15, 1963, was previously submitted by Consumers. That report covered experience from September 27, 1962, (initial criticality) until April 17, 1963. The current application contains operating information from April 17 through November 9, 1963.
The power operation test program is described in the final hazards summary report and outlined in the Technical Specifications. Results of the testing
~
conducted in accordance with this program indicate,.with two exceptions, no substantial variance from predictions with regard to thermal, hydraulic, and nucicar characteristics of the facility.
a
- m..
...-.m o,
,,,,.2-,.,-
..-r.,-%,..,,n
s, The first significant variation involved the neutron-source-startup-chamber georetry. To obtain the required three counts per second on the count rate meters the operating sources had to be positioned opposite the startup chambers.
This was authorized by Change No. I to the Technical Specifications dated November 28, 1962, which permitted the proposed repositioning but required that any one percent increase in reactivity above an estimated K-effective of 0.975 rust result in an increase of ten percent in neutron count. The-second variation was that the control blades were of higher reactivity worth than expected.
Because of their design, however, they were easily modified to reduce the relative worth of the individual blades within limits described in the FIral 1,);ctdsSummaryReport. This modification was authorized by Change No. 2 to I
the Technical Specifications dated January ~29, 1963.
A comparison of the calculated and experimentally determined principal thermal, hydraulic, and nuclear parameters of the reactor, with the initial core loading (no development fuel), is presented below Parameter Calculated Determined
- Total Peaking Factor 4.1 3.58 Max. Heat Flux @ 157 Mw(t), "
5 5
2 Btu /(hr)(ft )
3.62 x 10 3.15 x 10 Min. Burnout Ratio at Overpower 1.92 2.25 Specific Power, Fcted Conditions, Kw/f t 10.7 9.5 Total Recirc:lation Flow Rate, 6
6 lb/hr 12.4 x 10 13.1 x 10 The calculated and determined reactivity coefficients are also in good agreement.
[1rnt Operating History The operating history reflects some design deficiencies of the plant which have been or are being corrected. These include the followings (1) Control rod drive mechanism malfunctions.
(2) Reactor inlet diffuser design.
(3) Nuclear and process instrumentation problems.
With developmental fuel in the reactor slightly less conservative values were observed.
l i _
During the initial operations there have been three control rods which drif ted out of the core, and four rods that were stuck in the fully inserted position and could not be withdrawn. The drifting was attributed to resins and other foreign material in the collet assembly of the mechanism which held the assembly in the unlocked position. The primary system has since been cleaned and there have been no recent cases of rods drifting from the core.
It is our under-standing that Consumers is considering a modification of the collet assemblies to make such malfunctions less likely to occur, but that no action on the modifications has been taken at this time. It is our opinion that until the existing apparent deficiency has been corrected, close surveillance of the drive system by operating personnel should be continued.
The inlet flow diffusers in the reactor vessel were originally of a simple deflector baffle design. Twice, as a result of excessive vibration (ete 4.5 g accelerations) on one empty fuel channel on the core periphery, the four studs that attach the Zircaloy channel to its stainless steel support tube loosened and fell to the lower part of the vessel. These bolts lodged between the lower grid plate and the index tube of some of the drives preventing their withdrawal (as described above). The diffusers were modified, the channels were stiffened, guards were placed around the holes in the core support plate thct accommodate the drives, and a special keeper was welded to the studs.
Subscquent flow tests indicate that the vibration acceleration has been reduced well below 1 g, which is considered acceptable.
On three recent occasions both safety channels have indicated a scram (by proper annunciation) but no scram was initiated. Also it was not possible to tell which sensor caused the trip. The trouble was postulated as vibra-tion on a remote panel (located in the containment building on the floor over the main steam line) and the offending sensors were believed to be those indicating steam drum low water level. General Electric has conducted tests on the system and has concluded that the vibration is of sufficiently short duration to trip the annunciator relays but not the actual scram relays. The second instrumentation problem involved the failure of two transistors in the safety system scram logic circuits, which ney have been of non-fail safe nature, although other (redundant) transistors in the circuit would have had to fall to negate the system. Consumers corrected the deficiencies and submitted a report on its instrumentation problens and the corrective action taken by letter dated December 20, 1963, which is presently being evaluated by the Regulatory Staff.
Results of Phase I - Research and Development Program Significant experience was gained under Phase I of the Research and Develop-ment Program in the following areast (1) Irradiation of thin stainless-steel-clad fuel bundles in VBWR (2) Operation at pressures between 800'-1500 psig
(3) Utilizatioa of instrumented fuel bundles (4) Stability tests (5) -Utilization of the scheduling computer Experience with thin stainless-steel-cladding has shown that early cladding failure can be expected. Several Big Rock prototype development fuel bundles placed in VBWR have failed. The Phase I development bundles, eight in number, have been in use in the Big Rock reactor and present plans are to leave them in during Phase II. However, it is planned to test Inconel and Incoloy as cladding material in Big Rock, and 12 of the 15 Phase II fuel bundles will be clad with the'se materials, while the remaining three will be clad with Zircaloy.
The operating parameters of pressure, core inlet subcooling, and recircula-tion flow rate have been experimentally examined at Big Rock over wide ranges. The response of the reactor system to variation of these parameters appears to be well understood.
The four instrumented fuel bundles, used in connection with Phase I have not been entirely satisfactory. Failures of the flow meters and ion chambers have resulted in relatively little experimental data. However, General Electric has redesigned these probes and intends to use the same bundles during Phase II.
It is expected that data on flow, flux, and temperature will be obtained for Phase II.
Rod o cillator tests, performed during Phase I, indicate that the reactor has a large margin of stability. Predictions indicate no stability problems during Phase II operation at 240 Mw(t), and further tests will be undertaken to verify stability at this power level.
The scheduling computer has been installed and operated successfully. Con-sid.erable data is available to the operator through use of the computer, both on a scheduled basis and on-demand. Most of the significant reactor operating parameters are recorded; nuclean thermal, and hydraulic calculations are made; and, in addition, most of the alarm functions are reccrded and umy be readout on-demand. The computer provides no control function to the reactor, but data is available directly to the operators by printer readout in the control room.
Prnposed Escalation to 240 Mw(t)
The escalation to 240 Mw(t) is directly related to Phase II of the Research and Development Program since the Phase II fuel bundles will be in the 84 bundle core when the reactor is first taken above 157 Mw(t). A planned i
I e
~.
- b',
i f
" stepwise approach" to 240 Mw(t) is outlined in the application. Testing involving (1) power calibration of the nuclear instruments, (2) calibration of the in-core ion chambers, (3) radiation surveys, (4) radiochemical analyses, (5) bypass valve tests, and (6) initial pressure regulator tests are planned at l20 Mw(t) steps (approximately). A comparison of some of the current and proposed operating parameters are listed below:
Parameter Current Limit Proposed Limit Power level 157 Mw(t) 240 Mw(t) 2 Heat flux, overpower 510,000 Btu /(hr)(ft )
530,000 Btu /(hr)(ft )
@ 127% rated power
@ 122% rated power Burnout correlation APED 3892 APED 3892 Minimum specified 1.5 1.5 burnout ratio Specific power, rated, Kw/f t 13.1 14.2 Power density 45 Kw/ liter @ l57 Mw(t) 45 Kw/ liter @ 240 Mw(t)
Reactor pressure 800-1500, nominally Nominally 1250 psia 1000 psia The safety analysis section of the Final Hazards Summ.ary report for this f acility presented accident analyses based both upon 157 Mw(t) and 240 Mw(t) operation. A review of these analyses indicates that the consequences of the major accident considerations at the proposed higher power level, including the maximum credible accident, remain unchanged. The application indicates that the pertinent analyses of off-standard conditions, equipment malfunctions, and operator errors have been re-evaluated using performance data obtained from the power operation test program and Phase I testing.
Results of this re-analysis, which are not significantly different from results of previous analyses, are presented in tabular form in the application.
Research and Development Program,. Phase II Several aspects of Phase II of the Research and Development Program, in addition to operation at 240 Mw(t), are of particular interest. These are:
(1) Use of Inconel and Incoloy cladding (2) Testing at 60 Kw/ liter (3) Reactor control by varying recirculation flow rate (4) Extensive use of the scheduling computer (5) Power run to achieve 15,000 Mw-days / ton burnup
t Of the 15 Phase II development fuel bundles, three will be clad with Zircaloy-II (30 mils), three with Inconel (19 mils), three with Incoloy (19 mils), and six with Incoloy (11 mils). General Electric has had some experience with the use of Inconel and Incoloy as a fuel clad material in its superheat development work. Data thus far obtained indicate that these high nickel steels are more corrosion resistant and are less affected by irradiation than stainless steel and Elycaloy. In the General Electric tests to date there have been no corrosion failures of Inconel or Incoloy clad fuel rods principally because the threshold-to-failure, which is apparently well above that of stainless steel has not yet been reached.
After the testing at 240 Mw(t) (45 kw/ liter and 84 fuel bundles) a series of tests at 157 Mw(t) (60 kw/ liter and 41 fuel bundled will be conducted.
Stability data obtained at Big Rock Point thus far indicates that the rargin of stability is apparently more sensitive to core size than to power density. Calculations indicate that the operation at 60 kw/ liter (41 bundle core) will be more stable than operations at 45 kw/ liter (84 bundle core).
Tests to date of reactor control by varying recirculation flow rates (use of butterfly valves in the recirculation lines) proved inconclusive, because limits on valve travel were reached well before limits on reactivity change.
General Electric intends to conduct further tests during Phase II to determine the ' extent that such control is feasible.
Extensive use of the scheduling computer is planned for Phase II..Ghiculation of west effective control rod patterns will be undertaken and will provide guidance for rod programming. Control rod scheduling and positioning by the computer (i.e., computer control of the reactor), has not been proposed for Phase II.
The four and one-half year Research and Development program extends from the criticality date of September 27, 1962. The dynamic testing for Phase II described in the application is expected to last about six months. The balance of the program period, t,roximately two and one-half years, will be devoted to achieving a burnup of the Phase II development fuel of 15,000 megawatt days / ton.
Referral to ACRS The application from Cot sumers was referred to the Advisory Committee on Reactor Safeguards and considered at its January 1964 meeting. The Committee stated, in its report of January 17, 1964, that in its opinion the reactor can be operated with forced circulation in Phase II within the limitations proposed in the application amendment without undue hazard to the health and safety of the public. The Technical Specifications will preclude natural circulation flow tests until such times as evaluation of such teets has been received by the Regulatory Staff and the ACRS and the tests have been authorized according to the change procedures outlined in 10 CFR 50.59.
s
i Technical Specifications
/.ppropriate for a full term license, and required for operation at 240 Hw(t) and conduct of Phase II of the Research and Development Program, are several changes to the Technical Specifications to License No. DPR-6.
On the basis of inf ormation presented in Amendment No.15 and several discussions with the.
applicant, changes will be made in the following sections of the Technical Specifications:
(1) Section 3.7 - Containment testing, page 6 (2) Section 4.1.2(a) - Recirculation valves, footnote, page 10 (3) Section 4.1.2(b) - Conductivity operating limit, page 13 (4) Section 5.1.1 - Principal core materials, page 18 (5) Section 5.1.2 - Revised core drawing, page 20 4
(6) Section 5.1.5 - Core Composition, page 23 (7) Section 5.1.6(a) - Initial start-up sources, page 26 (8) Section 5.1.6(b) - Operating sources, page 26 (9) Section 5.2.1 - Principal calculated characteristics, 27, 28 (10) Section 5.2.2 - Calculated nuclear characteristics, page 28-31 (11)
Section 5.3.1(a) - Reactor power level, refueling, page 31 (12) Section 5.3.1(b) - Reactor operation, pages 31, 32 (13) Section 5.3.3 - Liquid poison system, page 36 (14) Section 6.1.2 - Reactor safety system, page 40 (15)
Section 6.1.5(f) - Operating requirements, page 45 (16) Section 7.6.2(e) - Cold startup, page 66 (17) Section 7.9 - Operational testing, page 71 (18) Section 8.0 - Research and Development Program, Phase II; pages 72-82
e t
(
(
g t.!,
. (1). Section 3.7 The'present Techn'ical Specifications require that, with regard to containment leak rate testing, "a procedure coverins uture inspection and method'of testing shall be submitted for AEC appro's1 prior to the elapse of 18 months following issuance of the provisional operating license."
. Ccnsiderable reliance is placed upon the containment building and its related systers to attenuate the consequences of a fission product release accident, such as those considered in Consumers' previous applications for a pro-visional operating license.
In view of this we believe that periodic
-integral Icak rate testing supplemented with more frequent leak detecticn tests of the components of the containment system are essential from the standpoint of nuclear safety to provide some degree of assurance that.ths integrity of the containment-is being reintained. Such testing consideratiens i -
are required particularly when the service life of such a " final barrier of containrent of fission products" is contemplated being extended over a forty year period.
Therefor.
a result of our review of the applicant's proposals, subsequent discussio.
i the applicant, and based upon our evaluation of the requirc-mants for i_
sting of the Big Rcck Point containment building, we believe the following eccifications should be incorporated into the Operating License Technical Specifications:
"3.7 CON!AINMENT SPHERE LEAKAGE TESTING 1
For the purpose of this specification, leakage rate is defined as the percent of the contained atmosphere (weight basis) which escapes
- r i
dsy (24 hrs) under the defined pressure conditions through any leaks in the containment _ boundary and all isolation valves and their associated piping.
Iht. nnxinem allowable integrated leakage rate shall not exceed 0.5%/ day of the containnent atrosphere (weight basis) at the design pressure of 27 psig. The procedure for containment sphere leakage testing shall be:
(a) At least once every 6 months, the personnel lock, the equipment lock and the sphere scpply-and-exhaust ventilation valves shall be pressurized, with air to 20 psig, to test their leak tight-ness. The sum of leakage rates from these valves and locks shall be less than 0.25% per day of the containment atmosphere (weight basis) at 20 psig.
(b) At least once every 12 months, the following valves shall be t ested for operability from both the manual and automatic modes of operation and, at the sace tine, shall be tested for-i y
e
g e
. leak tightness by means of a pressure test utilizing air or the normal working fluid at a pressure not less than 20 psig.
Fbin Steam Isolation (M0 7050)
Main Steam Drain (MO 7065)
Cleanup System Resin Sluice (CV 4091)
Reactor & Fuel Pit Drain Isolation (CV 4027 - CV 4117)
Reactor Enclosure Clean Sump Isolation (CV 4031 - CV ^103)
Reactor Enclosure Dirty Sump Isolation (CV 4035 - CV 4103)
All significant leaks revealed by these tests shall require repair of valve seals and retests.
Automatic controls and instrumentation associated with these valves shall be tested at approximately quarterly intervals; these tests may be conducted with simulated sigrals ur in such other manner as to obviate plant shutdown.
w (c) At least once every 12 nonths the following shall be visually examined for evidence of corrosion, cracking or deterioration:
All Electrical and Accessible Piping Penetration Nipple Welds All Accessible Piping Welds to Nipples All Expansion Joints and Welds on Expansion Joints Potting Compound in All Electrical Penetrations Insulation at piping penetration welds shall be removed to permit visual examination.
The probable cause of any significant corrosion, cracking or deterioration by such visual examination shall be determined, and evaluated in terms of likelihood of recurrence and probable ef-feet upon other containment sphere penetration components. An l
individual component leak detection test shall be perforned at 10 psig air on the faulty component prior to its repair or modifi-cation. The faulty component, and other components if necessary, shall be repaired or modified, and an individual component leak detection test perforned at 10 psig air upon each repaired or modified component. All components so repaired or modified shall be 'fisually re-examined at appropriate intervals, but not less frequently than once every six months, until the adequacy of annual visual inspection is reestablished to the operator's satisfaction.
Af ter cutting into the sphere or its components, or any dis-assembly of components that would af fect sphere integrity, an
/
J t.
ndividual component leakage rate or an integrated leakage rate itest, whichever is deemed more appropriate by the operator, shall be performed, with air at a pressure not less than 10 psig.
in lied It shall be permissible to employ a leak detection test of _the above for insuring containment integrity following dis-
- assembly of ~ the emergency condenser or the gasketed, bolted closure of the coaxial catie electrical penetrations.
The individual component leakage rate determined 'from the above tests when combined with the previously measured integrated leakage rate shall yield an.overall leakage rate not greater than 0.5%/ day at 27.psig.
least (d)- A containment sphere integrated Icakage rate test of at less 24-heurs duration shall be performed at a pressure not than 10 psig. Although routine maintenance rey be performed, repairs to items listed in 3.7(a), (b), and (c) shall not be made immediately prior to or during the test.
The accuracy of the leakage rate measuring system shall be verified (1)'by superimposing a controlled leakage rate equivalent to the allowable leakage rate at the test pressure (measured through a-gas flow meter) upon the existing leakage rate and continuing the test a sufficient period of time to neasure the composite leakage, or (2) by other reans of equivalent accuracy.
If the Icakage is in excess of 0.57./ day of contained atmosphere i
(weight basis) at the design pressure (27 psig) or extrapolated to the design pressure, repairs shall be made and the leakage rate test repeated until the 0.5% per day of contained atmosphere (weight basis) specification is net.
An integrated Icakage rate test shall be conducted on the contain-j
- ~
(e) nent sphere within 90 days af ter the effect;ve date of the full-term operating licensc, in accordance with the procedures set forth in subsection 3.7(d) above. On the basis of the results of such test, and other available relevant data, the operator shall, prior to 1966, propose a frequency of integrated leakage -rate t(sting covering the remaining term of the full-term operating license. If the proposed frequency is acceptable to the Division of Licensing and Regulation, the provisions of this Section 3.7 l-shall be amended accordingly. Othe rwise, such integrated - lea'. ge
~-
rate testing shall be repeated in 1966 and once every two years thereaf ter pending agreement upon a different frequency.
i
~-
t
-a w
y
+
i a
(fh' If the results of an integrated leakage rate test show' leakage above. specification, the required repairs shall be performed.
ine integrated leakage rate test shall then be repeated until the -0.5% per dey (by weight) specification is met and another integrated leakage-rate test shall be performed within 1 year; except that.if *:... excessive leakage rate is attributed solely to the containment components specified under' Section 3.7(a) the integrated l leakage rate test need not be repeated within one year, but the frequency of testing required by Section 3.7(a) shall be doubled for that period of one year.
(g) All leakage rates determined by tests at pressure less than 27 psig shall be corrected by the following extrapolation factor to establish the leakage rate at design pressure:
L 2
u e
.P
,g t
L t
P
-1 a
t i
where L,
= % leakage rate, at extrapolated pressure L
= % measured leakage rate, at air test pressure P
= extrapolated pressure, atmospheres, absolute e
P
= test pressure, atmospheres, absolute
= test viscosity of air, at test pressure and au temperature u
= viscosity of air-steam mixture at pressure a
and temperature of the accident condition."
t 12 -
(2) Section 4.1.2(a)
~
The limit of operation of the butterfly valves has been specified as 450 from the full-open position. Tests conducted during Phase I of the Research and Development Program have shown that such a static -
limitation does not allow much variation in the recirculation flow rate.
To allow such variable flow testing, within the core operating limita-tions set forth in Section 5.3, Consumers has proposed a dynamic specifi-cation limiting operation of the valves such-that a 50% reduction in flow from the full-flow condition be possible. Since this was the intent of the original specification, we believe it appropriate to change the foot-note on page ten of the Technical Specifications to the following:
"
- The limit of operation of each butterfly valve shall be adjusted to produce a 50% reduction in flow from the full-flow condition with the corresponding pump in operation."
(3) Section 4.1.2(b)
One of the operating limits of the primary cooland recirculation system, involving the maximum conductivity of the primary water, is, in our opinion, somewhat open to question as to interpretation. The specification states that 5 micromhos/cm may be exceeded during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a startup from cold shutdown. We believe that a maximum transient limit of 10 micromhos/cm should be incorporated and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit further clarified. This specification, page 13, should be revised as follows:
" Conductivity (Micromho/cm)
Maximum 5
I Maximum transient
- 10
"
- Conductivity is expected to increase temporarily after startups
,Twomicold ' Ahutdown. The maxinum transient value here stated is the maximum permissible and applies only to the period subsequent to a cold shutdown between criticality cad 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 20% rated power."
(4)' Section 5.1.1 Reflecting the use of Incoloy 800, Inconel 600, and Zircaloy-II as fuel cladding, Section 5.1.1, page 1% should be changed to the followingt
" Principal Core Materials Fuel (Sintered Pellets or UO Compressed Powder) 2 Moderator and Reflector Light Water 9
f i
13 -
" Structural Components (Fuel Cladding 304 SS and Incoloy 800 Wall, in addition to 304 SS and Incoloy 800, include Zr-2 and Inconel 600)
Flow Channels Zircaloy-II and 304 SS Control Rods B4C Filled 304 SS Tubes and 304 SS Rods in Crucifccm Shaped 304 SS Sheaths" (5) Section 5.1.2 Th'e core configuration drawing presented in Consumer's application
( Amendment No.15) reflects the fully loaded 84 bundle core and should replace the 56 bundle core drawing on page 20 of the Technical Specifi-cations.
(6) Section 5.1.5 To be consistent with the core composition changes described in Consumer's Amendment No. 14 and reflecting the staff's position on the acceptability of such changes, Section 5.1.5, pages 23, 25, and 26 of the Technical Specifications should be revised as follows:
" General Core Composition The data in this section present general design features of the original and research and development fuel that shall make up the
-physical composition of the core.
(a) Enrichment of Fuel, approxi ate "ei ht percent U-235 from 2.6 to 4.5, inclusive.
(b) General Core Data Namber of Fuel Bundles in Core 84 Total Weight UO in 84 Bundles, Pounds 29,300 2
Moderator to Fuel Volume Ratio 2.65 Equivalent Core Diameter, Inches 76.54 (c) Fuel Bundles The 3eaeral dimensions and configuration of the two types of fuel bundles shall be as shown on pages 24 and 73 of these specifications.
Principal design fectures shall be essentially 2s follows:
1 1-w
~
.5 I
- 14 Resenreh and General original Fuel Developenent ruei Ocometry, Fuel Rod Array 12 x 12 11 x 11 Rod Pit;h, Inches 0.533 0.580 jjb{f Standard Fuel Kode per Bundle 132 109 Special Fuel Rods per Bundle 12 (4 Special Fuel 12 Rods at Bundle Corners are Segmented)
Spacers per Assembly 3
7 Fuel Red Cladding Ibterial 304 SS 304 SS, Zr-2, Inconel 600 and/or Incolo" ECO Standard Rod Tube Wall, Inches 0.019 0.010 to 0.030,
/
Inclusive Special Rod Tube Wall, Inches 0.031 0.010 to 0.030, Inclusive Fuel Rods Standard Diameter, Inches 0.388 0.425 Special Rod Diameter, Inches 0.350 0.320 UO Density, Percent Theoreti-2 cal 94 + 1 90 to 95, Inclusive Active Fuel Length, Inches Standard 70 68 to 70. Inclusive Corner 59 Fill Gas Helium Helium (d) Channel s Number of 304 SS and/or Zircaloy II 88 Wall Thickness, Inches 304 SS 0.075 Zircaloy II 0.100 I
m.
\\
Inside Width, Inchest 304 SS 6.57 Zircaloy II 6.54 Length, Inches:
304 SS 79-5/8 Zircaloy II 79-3/4 (e) "Totsi VelRht Supported by Core Support Platet 84 Fuel Bundles @ 420 Lb/ Bundle, Lb 35,280 88 Support-Tube-and Channel Assemblies
@ 100 Lb/ Assembly, Lb 8,800 86 Orifices @ 10/ Orifice, Lb 860 2 Channel Plugs @ 10 Lb/ Plug, Lb 20 1 Flow Distributor Assembly, Lb 2.500 Total Weight, Lb 47,460 "
(7) Section 5.1.6(n)
Consumers -has indicated that the five curie plutonium-beryllium source used for the initial startup and low-power testing of the Big Rock Point Reactor
.will no longer be required. Accordingly, the present Section 5.1.6(a), page 26, should be deleted.
(8) Section 5.1.6(b)
TE.*chnnection with the above change, Section 5.1.6(b), page 26 should be redesignated Section 5.1.6.
Additionally, the location specification should be changed and a footnote added as follows:
" Location * - The sources shall be placed in core positions 02-59 and 09-52 as shown on page 20.
If the required three counts per second count rate with a 3:1 signal-to-noise ratio cannot be obtained with the sources in these locations, it shall be permissible to use post-tions 02-52 and 09-59.
"* With in-vessel low level neutron detectors in service, one operating source may be temporarily relocated as the operator deems appropriate."
The operational requirements for the sources in these locations will be further discussed in item (16) and outlined in Section 7.6.2(e) of Wha i
Technical Specifications.
8 l (9) Section 5.2.1 The principal calculated thermal and hydraulic characteristics of the core, described in Amendment No. 14 for the 84-bundle core loading should be incorporated in the Technical Specifications, pages 27 and 28 as follows:
"5.2.1 Principal Calculated Thermal and Hydraulic Characteristics of the 84-Bundle Core Loadina (a) Core Power at Rated Steam Flow, Mwt 240 (b) Peaking Factors (To be applied to Heat Flux):
(1)
Overall at Rated Power (Includes Gross and Local)*
2.83 (ii)
Overpower (Steady Stete and Transient Effects) 1.22 (iii) Total (Product of (1) and (ii) 3.45 (c) Heat Flux and Fuel Center Temperatures Average Maximum Maximum @ 122%
@ 240 Mwt
@ 240 Mwt of 240 Mwt Fuel Jacket Heat Flux 2
(Btu /Hr-Ft )
116,500 329,000 402,000 Cladding 620 720 760 Fuel Center, Temperature, OF 1,380 3,350 4,0 50 Fuel Rod Power, Kw/Ft 3.8 10.7 13.1 (d) Burnout Ratio, Minimum at Overpower 1.8 (e)
Maximum Fuel Cladding Stress, Psi 72,200 (f) Average Core Power Density at 2,40.,Mwt,Kw/L 4pe (g) Stability Margin at 240 Mwt, 1485.Psig Degrees Phase Margin 28
(h) Total Recirculating Flow Rate, Normal for 6
240 Mut, 1235 Psig, Lb/Hr 12.5 x 10 (i) Reactor Core Flow Rate Percent of Total Recirculation Flow Rate 98 l
Phase 1 Developnent Power Fuel l
i
(j) Core Inlet Conditions Inlet Velocity Maximum, Ft/Sec 4.5 Minimum,Ft/Sec 2.7 Inlet Subcooling, Btu /Lb 19.5 (k) Reactor Core Pressure Drop, Normal for 240 Mwt, 1235 Psig 5.7 (1) Reactor Core Maximum Exit Bulk Temperature, OF 572 (m) Steam Volume Fractions Average Core Exit Fraction 0.55 Maximum Channel Exit Fraction 0.63 Average Fraction over Core Length 0.31" (10) Section 5.2 1 In accordance with the above change, the principal nuclear characteristics, Section 5.2.2, pages 28-31, should be revised as follows:
"$.2.2 Principal Calculated Nuclear Characteristics of the 84-Bundle Core (3) Temperature and Void Coefficients The following temperature and void coefficients have been calculated for the 84-bundle core. These values are based on calculated characteristics of the developmental bundles to be inserted together with experimental data obtained relative to the characteristics of the original fuel. Data relating to individual research and development fuel types are included in Section 8.1.1.
~
Moderetor Temperature Coefficient (/1k gg/k gg per F) e e
6F F 550 F Start of Cycle
+0.2 x 10-4
-1.5 x 10-4 End of Cycle
+0.6 x 10
-0.9 x 10~'
Void Coefftetent (fik,gg/k,gg per Unit Void Within the Channel)
, 68 F 550 F Start of Cycle
-0.20
-0.20 End of Cycle
-0.10
-0.10
s.
i
+
18 -
~#
l (b) ponnier coefficient (6 k, /k,per F)
Fuel Temperature F Moderator F (A km /km / F
-5 68 68, 07. Voids
-1.47 x 10
-5 1323 550, 07. Voids
-1.03 x 10
-5 1323 550, 207. Voids
-1.15 x 10 (c) Reactivity Balance ok gg Temperature 0.025 Voids 0.030 Xenon and Samarium 0.025 Fuel Depletion and Maneuvering 0.050*
Total Ak Required 0.130 (d) Multiplication Factor. Cold Condi- "k,gg tion All Rods In, 18 SS Channels 0.95 All Rods Out, 18 SS Channels 1.13 31 Rods In, 18 SS Channels 0.98 (e) Maximum Control Rod Worth Ak,gg/k gg Cold 0.039 550 F, No Voids 0.042 550 F, 207. Voids 0.030 (f) Maximum Reactivity Addition Rete (Based on Withdrawal of the Rod Worth 0.042 Ak gg/k gg at the e
e Maximum Withdrawal Rate of three Inches per Second),Ak gg/keff e
per Second 0.0052 The effect of removing SS channels is not included in this number.
-Removal of 18 SS channels would add 0.050 dk,gg.
(g) Worth of Liould Poison eskegg, Normal Water Level in Reactor and
-0.30 680F, 2000 Ppm Boron lik ff, Normal Water Level in Reactor and 5506,F, 1300 Ppm Boron
-0.19 Maximum Time to Bring Ron: tor Suberitical, 5
Hot, Hinutes Time Required for Poison to Reach Core After Activation of the System, Seconds 35 (h) Expected Fuel Burnup Average Mwd / Ton of Contained U Original and Phase I Fuel Types 10,000 Phase II Fuel Types 15,000" (11) Section 5.3.1(a)
Since use of the plutonium-beryllium source has been discontinued, Section 5.3.1(a), page 31, may now be revised as follows:
"5.3.1 Reactor Power Level (a) Refueling The reactor power shall be limited to 1.0 Mut, exclusive of core decay heat, during operations with the reactor vessel closures open."
(12) Section 5.3.1(b)
Consistent with the remarks in this evaluation on maximum heat flux and maximum power density, Section 5.3.1(b), pages 31-32, the core operating limitations should be amended as follows:
"(b) Reactor Operation Reactor operation shall be so limited as to be consistent with the rest conservative of the following:
- Minimum Core Burnout Ratio at Overpower 1.5 Transient Minimum Burnout Ratio in Event of
'- 1.5 Loss of Recirculation Pumps from Rated Power 2
Maximum Heat Flux at Overpower, Btu /Hr-Ft 530,000
- Based on correlation given in " Burnout Limit Curves for Boiling Water Reactors" by E. Janssen and S. Levy, April 14, 1962 - APED 3892.
1 1
20 -
Maximum Steady State Rod Power, Mw/Ft 14.2 Maximum Steady State Heat Flux, Btu /Hr-ft 434,000 2
Maximum Fuel Rod Power at Overpower, Kw/Ft 17.2 Stability Criterion: Maximum Measured Zero-to-Peak Flux Amplitude, Percent of Average 20 Operating Flux Maximum. Steady State Power Level, Mst 240 Maximum Value of Average Core Power Density 60
@ 157 N (t), Kw/L Maximum Value of Average Core Power Density
@ 240 N(t), Kw/L 46 Maximum Reactor Pressure During Power Operation, 1485 Psig 6
Minimum Lecirculation Flow Rate, Lb/Hr 6 x 10 Maximum Nd/ Ton of Contained Uranium for an)
Individual Bundle 235,000 Rate of change of reactor power during power operation:
Control rod withdrawal during power operation shall be such that the average rate of change of reactor power is less than 50 Mwt per minute when power is less than 120 Mwt, less than 20 h t per minute when power is between 120 and 200 Mwt, and 10 ht per minute when power is between 200 and 240 Nt."
(13) Section 5.3.3 The present Technical Specifications require, with respect to test firing of the liquid poison system valve squibs, that "af ter 12 months' operation, 3 squibs (one from each operating valve group) shall be removed and test-fired." Such testing was satisfactorily performed. Consumers has now pro-posed that the last sentence of Section 5.3.3, page 36, be amended as follows:
"One squib (from diff.: =,t operating valve groups and valve position each time, in N ation) shall be removed and test fired at intervals of not less than 12 months."
We agree that this is an acceptable specification.
(14) Section 6.1.2 In accordance with item 2 the scram setting listed in Section 6.1.2, page 40, for the condition of recirculation line valves closed should be changed to read as follows:
"Approximately 107. of full simultaneous closure of both discharge or both suction valves or simultanews closure of the butterfly I
valves to the positions comparable to a 557. decrease in flow from full flow or any combination of two of these valves, one in each l
loop."
l
!~
\\
g.
v a
i
(
(15) Section 6.1.5(f)
To make the requirement for operating in-core flux monitors at 240 Mw(t) consistent with the present requirement given in Section 6.1.5(f), page 45, the value of "120 Mwt" be changed to "180 Hwt."
(16) Section 7.6.7(e)
Section 7.6.2(e) requires that "when k-effective is greater than 0.975, an increase of a minimum of ten percent shall be observed on at least one neutron monitoring channel for any estimated one percent increase in k-effective." This specification was developed as Change No. I to the
N Technical Specifications dated November 28, 1962, as described in the related hazards analysis. The purpose of the change was to provide operational compensation for a poor neutron-source-startup-chamber geometry. In view of the change discussed in item 8, a specification generally applicable to both proposed new source positions should be applied by amending the last sen-tence of Section 7.6.2(e), page 66, to the following:
vity (Q)g k-ef f ective is greater than 0.970, any incre ase in reacti-
"Wheneve greater than 0.001 shall result in a fractional 'ncrease reactivity ( AE( 21 c _) of not less than 50% of the fractional change in in count rate -
)!
2.
The following footnote should be added to this page.
1-k,gg
, q k,gg (17) Section 7.9 Consistent with the change described in item (13) the frequency of testing notation for the liquid poison system listed in Section 7.9, page 71, should be changed to the following:
" System or Function Undergoing Test Freauency of Routine Tests Liquid Poison System Component Two Months or Less During Power Operability.
Operation. One Squib Test-Fired Each 12 Months."
In addition, the phrase "during the period of the provisional operating license" should be deleted 'f rom the footnote on page 71 and f rom the following referenced sections of the Technical Specifications which relate to the footnote:
1.
Section 3.5.2(c) - page 6 2.
Section 5.3.2(a) - page 33
i s
3.
Section 6.1.5(a), (b) - page 44 (footnote) 4.
Section 6.2.2
- page.46 (18) Section 8.0 Section 8.0 of the Technical Specifications is devoted to Phase I of the Research and Development Program. To reflect the precepts of the proposed Phase II, this section should be revised Ag,the following manner:
(a) In Section 8.0, Page 72, change the first paragraph to read:
"After completion of Phase I research and development testing at the Big Rock Point Nuclear Plant, Phase II of the Research and Development Program will be initiated. All normal operating limitations will be observed, except as specified in Section 8.3.
Delete 'and the installation' from the first sentence of the second paragraph.
In the first line of the third paragraph, change ' Phase I" to Phase II.'"
(b) Change Section 8.1.1 entirely:
Development Fuel Desian F#1tures Fifteen developmental fuel bundles have been designed for use in the Big Rock Point reactor core at the commencement of Phase II.
Addi-tional developmental fuel bundles mayt be provided at a later date in the Phase 11 effort. Although the details of these additional fuel bundles may differ somewhat from those described here for the origi-nal 15 fuel bundles, their principal design features will be similar.
All fuel bundles are of a similar mechanical design typified by the bundle drawing in Figure 8.1.
The design of the Phase I and Phase II developmental fuel is such as to give nuclear characteristics and equivalent to the original core l-fuel. The principal difference between Phase I and Phase 11 fuel is in the higher reactivity of Phase II bundles to permit operation to a burnup of 15,000 mud /T. Four different cladding materials are employed in the fuel designs. Enrichments for these bundles have been chosen to yield essentially the same reactivity for each of the types.
These reactivities are:
l l
l w
m--
m m-e y
e---
m 7
- c
l k oo 20 C 1.32 2880 C 1.32 2880 C + 20% Void 1.30 Void Coefficient
-27 x 10-4 sik/% Void Moderator Coefficient
+ 0.1 x 10-4 0/C (Based on Critical Leakage)
The design of the research and development bundles shall be such that the fuel rods may be replaced with rods containing poison material to effect either power shaping or reactivity control. Such rods may be tested for power flattening as part of the Phase 11 Research and.
Development Program. The effect of these rods shall be to reduce the "k" values and to make the temperature coefficient more negative.
The principal design features of the Phase I and Phase II developmental fuel are essentially as follows:
Geometry - Fuel Rod Array 11 x 11 No. of Standard Rods 109 No. of Special Corner Rods 12 Standard Fuel Rod Diameter, Inches 0.425 Special Fuel Rod Diameter, Inches 0.320 Spacers per Bundle 7
Average Heat Flux @ 45 Kw/L, Btu /Hr-Ft 116,500 Hydraulic Diameter, Inches 0.58 Core Volume per Bundle, Liters 63 Phase I Phack II Clad Material 304 SS 304 SS Zr-2 Inconel Incolov Incolov No. of Bundles 4 (Powder) 4 3
3 3 6 (Powder)
Clad Thickness, Inches 0.010 0.010 0.030 0.019 0.019 0.011 Nominal Clad-UO2 Gap, Mils 0
5 7
6 6
0 Enrichment, Percent 2.7 2.7 2.8 4.5 4.1 3.3 U02 Density, Percent 91 95 95 95 95 90 i
Water to Fuel Ratio 2.4 2.4 2.96 2.63 2.63 2.4 2
76.5 76.5 76.0 74.6 74.6 76.5 Heat Transfer Area, Ft Active Fuel Length, Inches 70 70 69.3 68 68
'70 UO2 Veight/ Bundles, Lb 370 385 302 336 336 374 Max. Fuel Cladding Stress, 72,200 32,400 14,600 24,200 26,400 70,000 Psi
i 24 (c) Replace Figure 8.1, page 73, with Figure 1, Amendment No.14, dated November 15, 1963.
(d) Replace Figure 8.2, page 74, with the revised Figure 8.2 in Amendment No. 15, dated January 17, 1964.
(e) Change the first sentence of Section 8.1.2 on Page 75 to read:
"One or mere instrumented bundles will be utilized for selected tests during Phase II."
veletc the last sentence in Section 8.1.2.
(f) In Section 8.2.1, page 76, delete first paragraph and insert the following:
"The Phase 11 core performance, transient and stability testing will be conducted over a range of variables as indicated by the following:
Variable Range Core Size, No of Bundles 40 to 86 Reactor Power Level, Mwt Up to 240 Recirculation Flow Rate 6 x 106 lb/hr to Full 2-Pump Flow Reactor Pressure, Psia 800 to 1500 Core Inlet Subcooling Rated to Maximum" In the next to last paragraph on Page 76, change the first sentence to read:
" Recirculation flow rate shall be set by positioning each of the butterfly flow control valves at appropriate points between the
' full open' and the '50% of rated flow' positions."
In the last paragraph on Page 76, substitute " Phase II" for " Phase I" in the second 11 net and delete "157 Hwt core as well as the" in the third and fourth lines.
l (g) Change Section 8.2.2 on Page 77, in its entirety to read "In addition to the irradiation of developmental fuel, a sequence of operational tests to provide detailed knowledge of the steady state and transient thermal-hydraulic characteristics will be conducted."
g f
1.
Full Power Operation The plant power shall be increased to approximately 240 Mwt in approximately 20 Mwt steps from the 157 Mwt rated condition.
Satisfactory performance with respect to safety shall be established at each step before proceeding to the next step.
The test program shall ultimately demonstrate 75 Mwe operation at a thermal power and core configuration capable of meeting the established reactor operating limits. Tests and observations to be made at each of the incremental power increases shall be typified by the followings a.
Power Calibration of Nuclear Instruments - Picoamneter readings shall be compared with thermal power level calcu-lations.
b.
Calibration of the In-Core Ion Chamber System - Flux wire data whall be used to calibrate the in-core chamber readings.
c.
Radiation Surveys - Portable and fixed instruments shall be used to measure neutron and gamma radiation in various plant areas.
d.
Radiochemical Analysis - Grab samples from appropriate stacions shall be analyzed.
Bypass Valve Test - The characteristics of the bypass valve e.
system shall be rechecked. It is the intent of such testing to assure that the bypass valve control system will control the reactor vessel pressure adequately under manual and auto-matic control conditions.
f.
Initial Pressure Regulator Tests - The operation of the initial pressure regulator system shall be tested at various set points and its effect on reactor operation recorded.
2.
llinh Power Density Demonstration With a reduced size core, typified by Figure 8.2, a stepwise increase in power shall be made from approximately 115 Mwt to approximately 1
157 Mwt where the target power density of 60 Kw/L average is to be achieved. The approach to 157 Mwt from 115 Mwt shall be made in at least three approximately equal steps. Core performance evaluation shall be made at each step increase to assure the satisfaction of all operating limits such as burnout ratio, heat flux, kilowatts per foot, and stability.
e
. 3.
Performance Testing As a part of the Phase II developmental program, tests similar to Phase I tests shall be conducted at escalated conditions compatible l
with the new core operating levels. Wire irradiation and gamma scans shall be performed at appropriate points preceding or during a sequence of Phase II tests to provide the necessary core power distribution data and to assure that operational limits are satis-fled. In addition, shutdown margin checks, and temperature co-ef ficient measureuents shall be made whenever necessary or whenever significant. changes havs occurred in the core. New developmental fuel bundles shall be subjected to fuel verification testing.
The Phase II core performance, transient and stability testing shall be conducted over a range of variables as indicated in Section 8.2.1.
In addition, recirculation pump trip tests may be performed under selected conditions, shown by prior analysis to be within operating limits, in order to evaluate coast-down characteristics, power decay characteristics, etc.
The combinations of the variables shall always be chosen and analyzed to satisfy the reactor operating limits specified in Section 8.3.
Maxinum subcooling shall be that corresponding to full bypass of steam around the turbine. Butterfly flow control valves in the two recirculation lines shall be utilized to control flow to 50% of the full flow from a given pump.
Other variations which may be imposed during the testing sequence would involve alternate control rod programs and variations in core inlet orificing, including removal of orifices. Prior to execution of the individual tests chosen within the parametric range described in Section 8.2.1, all limiting conditions shall be established by thermal-hydraulic and stability evaluations.
Transient hydrodynamic performance and stability shall be studied by introducing controlled disturbances such as reactivity oscillation, pressure set point changes, and flow variations at selected reactor operating conditions. Stepwise increases in reactor power level and decreases in core flow rate shall be observed in each test to evaluate progress and safety.
4.
Developmental Test Secuence During and af ter the steps followed to achieve full power, the Phase II developmental test sequence shall be approximately as follows l
.+-J t
t
- a. ' Core size, 84 bundles.
b.
Increase power from 157 Mut to 240 Mwt stepwise at full flow and rated subcooling.
c.
At selected increasing power level steps, decrease flow from full two pump flow in a stepwise manner at rated subcooling.
d.
Repeat seleeced tests from "b" and "c" for a condition of tax 17.mn subcooling.
e.
Decresse core size to approximately 41 bundles.
f.
Increase power stepwise from approximately 60% of the ultimate power until that power level is attained.
g.
Repeat tests described in (c) above with the small core.
h.
Increase subcooling and repeat selected tests from (f) and (g) above.
(h) '6elete Section U.2.3 on Page 77 in its entirety since this test is not planned for Phase II.
(1) Change Section 8.2.4 on Page 78 in its entirety to read as follows, and redesignate as 8.2.3.
"8.2.3 Analyses of Typical Testst 4
Analyses of conditions anticipated at the full power and highest power density operations result in the expected core therwal and hydraulic characteristics f
given in Table 8.1.
All reactor operating limits are met under these conditions. Other Phase II development tests, though differing in specific variables which may result in limiting conditions, will comply with the operating limitations specified in Section 8.3."
(j) Delete Section 8.3 (c), on Page 81.
Redesignate Section 8.3 (d) on Page 81 as Section 8.3 (c). Changc 2
2 Btu /Hr-Ft " and change Btu /Hr-Ft " to read "530,000 "510,000 "16.6 Kw/Ft" to read "17.2 Kw/Ft."
Redesignate Section 8.3 (e) on Page 81 as Section 8.3 (d). Delete i
the second paragraph and summary table, and substitute the followings-I
The operating limits specified in Section 5.3.1 (b) shall l
apply to all test operations during Phase II of the Research and Development Program whether utilizing original core fuel, I
instrumented fuel, Phase I developmental fuel, or Phase II developmental fuel."
TABII; 8.1 Pl!ASE Il CALCULATED THElutAL-ilYDRAULIC OIARACTERISTICS Peck Thermal Power Peak Power Density Number of Fuel Bundles in Core 84 41 Reactor Pressure, Psia 125' 1250 Reactor Thermal Power, Fkt(a) 2e 157 Average Power Density, Kw/L 4L 60 Fuel Type Development Initial Development Initial Peaking Factors Local 1.407 1.407 1.418 1.418 Gross 2.01 1,823 1.943 1.762 Overpower 1.22 1.22 1.22 1.22 Total (Product) 3.45 3.13 3.36 3.05 2
lleat Flux, Btu /llr-Ft Core Average at Rated Power 116,500 116,500 157,110 157,110 Maximum at Rated Power 329,000 299,000 434,000 393,000 Maximum at 122% Overpower 402,000 364,600 528,000 478,900 ei Kw per Foot of Rod 4
bbximum at Rated Power 10.7 8.9 14.2 11.7
}
Maximum at 122% Overpower 13.1 10.9 17.2 14.2 i
Minimum Burnout Ratio t
Steady State Overpower 1.83 2.00 1.52 1.68 Loss of Pump 1.50 1.50 1.80 1.98 Total Recirculation Flow, Lb/llr x 106 12.5 12.5 9.86 9,86 Core Inlet Subcooling, Btu /Lb 19.5 19.5 19.5 19.5 (a) At natural circulation with the 84-bundle core, a rated power level of 190 Mwt is calculated to have a burnout ratio above 1.5 At natural circulation with the 41-bundle core, a rated power IcVel of 120 Mwt is calculated to have a burnout ratio above 1.5.
e b
^ ~
6 t
(k) Section 8.5, Page 82, Change Phase I to Phase II.
(1)
Insert Table 8.1 of Amendment 15 as page 79. The current page 79 is no longer applicable to the Research and Development Program.
Conclusions Based upon the considerations discussed in this report. including the results of the power operation test program, the fact that the escalation to 240 Mw(t) is predicated on the generally accepted principle of the " stepwise approach," and that Phase II of the Research and Development Program will be conducted as an extension of and be based upon Phase I, we have concluded thatt (1) The proposed testing for power escalation is sufficient in scope to determine the optimum and safe operating parameters of the plant, up to and including those at a power level of 240 Mw(t).
(2) There is reasonable assurance that the health and safety of the public will not be endangered by the proposed operations.
(3) The Technical Specifications should be revised as indicated in the previous section of this evaluation to accommodate the pro-posed operations.
Robert H. Bryan, Chief Research and Power Reactor Safety Branch Division of Licensing and Regulation Dates IU' I
i I
._.