ML19323G846
| ML19323G846 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 05/21/1980 |
| From: | Rickover H ENERGY, DEPT. OF |
| To: | Clint Jones PENNSYLVANIA, COMMONWEALTH OF |
| Shared Package | |
| ML19323G847 | List: |
| References | |
| NUDOCS 8006090098 | |
| Download: ML19323G846 (13) | |
Text
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48 s
9 Department of Energv Washington D.C. 20585 May 21,1980 Mr. Clifford Jones, Secretary Department of Environmental Resources Commonwealth of Pennsylvania Ninth Floor, Fulten Building Harrisburg, PA 17120
Dear Mr. Jones:
This is in reply to your letter dated May 8,1980, in which you requested additional infornation concerning Naval Reactors' plans to continue the operation of the Light Water Breeder Reactor (LWBR) Core in the Shippingport Atomic Power Station beyond the originally planned minimum lifetime of 18,000 EFPH.
You requested updated safety analysis information confirming that the core can continue to be operated safely and any other safety or environmental impact analysis applicable to continued operation and eventual decommissioning of the Shippingport Atomic Power Station.
On May 1,1980, the Advisory Committee on Reactor Safeguards (ACRS) reviewed plans for continued operation of the LWBR Core with representatives of the Division of Naval Reactors, the Westinghouse Bettis Atomic Power Laboratory, Duquesne Light Company, and the Nuclear Regulatory Commission (NRC) staff.
The ACRS concluded that it was acceptable to operate the LWBR Core in the Shippingport Atomic Power Station beyond 18,000 EFPH, as proposed.
I have enclosed for your information a copy of the information report to the NRC and ACRS concerning the extended core operation and correspondence from the NRC and the ACRS which indicate they consider the planned continued opera-tions to be acceptable. As indicated in my March 10, 1980, letter to you, releases of radioactivity from the Shippingport Station continue to be far below applicable standards as documented in the annual Environmental Reports and the conclusions contained in the LWBR Program Environmental Statement (ERDA 1541) remain valid for the planned continued operations of the LWBR Core.
Responsibility for the eventual decommissioning of the Shippingport Atomic Power Station folicwing removal of the LWBR core rests with the Deputy Assistant Secretary for Nuclear Waste Management of the Department of Energy. Studies are underway regarding decomissioning alternatives and a notice of intent to prepare a Draft Environmental Impact Statement on the de:ommissioning will be issued in the near future.
I have been assured that the State of Pennsylvania cocinents for the decomissioning Environ-THIS DOCUMENT CONTAINS P00R QUAUTY PAGES 8'006090
Mr. Clifford Jones, Commonwealth of PA 2
mental Impact Statement will be sought as was the case for the LWBR Program Environmental Impact Statements.
Sincerely, H. G.
ICK0V$R Deputy Assistant Secretary for Naval Reactors Attachments:
(a)
Information Report Concerning Extended Operation of the LWBR Core at Shippingport (b) Letter from H. R. Denton, NRC, to Adm. H. G.
Rickover, Naval Reactors, DOE;
Subject:
Continued Operation of Shippingport LWBR, dated April 30, 1980 (c) Letter from Dr. M. S. Plesset, ACRS, to Chairman John F. Ahearne, NRC, dated May 6, 1980;
Subject:
Extended Operation of Shippingport Light Water Breeder Reactor forsarded by letter from R. Fraley, Executive Director, ACRS to Adm. H. G. Rickover, dated May 6, 1980 Copy to:
G. W. Cunningham, Assistant Secretary for Nuclea" Energy, DOE S. Meyers, Deputy Assistant Secretary for Nuclear Waste Management, DOE Advisory Committee on Reactor Safeguards, NRC Office of Nuclear Reactor Regulations, NRC NRC Public Document Room B. F. Jones Memorial Library, Aliquippa, PA
- ~
INFORMATION REPORT CONCERNING EXTENDED OPERATION OF THE LWBR CORE AT SH!?PINGPORT I.
Introduction The objective of the Light Water Breeder Reactor (LWBR) program is to develop technology that would significantly improve the utilization of the nation's nuclear fuel resources employing the well-established water reactor technology.
To achieve this objective, work has been directed towards the analysis, design, component tests, fabrication, installation and operation of a water-cooled, thorium / uranium-233 fuel cycle breeder reactor core at the Shippingport Atomic Power Station.
In support of this operation, an Environmental Impact Statement and Safety Analysis Report (SAR) were prepared and submitted to appropriate government agencies for review. The Nuclear Regulatory Commission reviewed the Safety Analysis Report and the Environmental Impact Statement in 1976 and issued two Safety Evaluation Reports (NUREG 0083 and Supplement dated July 1976 and October 1976, respectively). The Advisory Committee or. Reactor Safeguards
,cencluded that LWBR Core operation in Shippingport was considered acceptable in its letter dated August 19, 1976. The Safety Analysis Ecport and Environmental Report were also reviewed in 1976-77 by the State of Pennsylvania Department of Environmental Resources. The SAR stated that it was planned to operate the LWBR core for a minimum of 18,000 EFPH. The LWBR core at Shippingport was taken critical on August 26, 1977 and reached full oower on September 21, 1977. The plant was released for routine comercial electric power generation on December 2, 1977. As of February 29, 1980, the core has achieved 17,274 critical hours out of a possible 22,027 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> with a cumulative depletion of 15,130 EFPH.
In parallel with ongoing operation of the core, additional analyses have been perfomed to detemine the full potential of the LWBR core. These analyses,
)
confirmed by satisfactory test data obtained at Shippingport, indicate the core has a potential reactivity lifetime of more than 18,000 EFPH while still re-1 taining a breeding capability.
It is planned at this time that power operations at Shippingport be extended through 24,000 EFPH.
In addition, analyses are in progress to determine the feasibility of operating to 30,000 EFPH (or beyond).
These extensions will significantly increase the information obtained from the LWBR program particularly with regard to the fuel element and nuclear perfomance technology of the thorium / uranium-233 fuel system. Continuing operations of the present core beyond 18,000 EFPH will provide important technology for evaluating the Light Water Breeder Reactor as a future energy resource for the nation.
The objectives of extended-life operation are to demonstrate improved uranium and thorium fuel utilization by producing additional electrical energy while maintaining breeding conditions in the core, to further verify the reactor design calculational methods for the LWBR fuel system and to gain additional operational experience with the UO -Th0 -based fuel system to higher irradiation exposures.
2 2
Extended core operation will be accomplished by reducing the maximum operating
. power, main coolant pressure, and operating temperature which will maintain adequate fuel element perfomance by reducing fuel and clad temperatures and Attachment (a) :: Sta:e of :A Letter NBI Log No. 0203-B0/COSIL
cladding stresses while retaining the breeding capability in the core. Similar extended core operations under reduced power and temperature conditions were perfomed with the previous two cores operated in the Shippingport Atomic Power Station.
II.
Perfomance to Date During operation to date, performance of the LWBR core has been closely monitored.
Core follow calculations using the nuclear design model have shown the core at full power to be slightly more reactive than nominal (by less than 0.5%
o t in reactivity). This is within the core themal and fuel element per-fomance assessments reported in the SAR which were conservatively based on power distributions associated with core reactivities varying i 1% A d in reactivity from nominal.
Periodic primary coolant chemistry analysis of radioactive iodine isotope concentrations and continuous delayed neutron monitoring have shown no indication of' fuel element cladding defects.
As an added test of fuel element perfomance 153 load follow cycles (from 90% of full power to between 60 and 35% power for periods of 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) have been included as part of regular power operations since initial operations
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of the LWB?. core.
Based on the actual power history, stress in the cladding of the fuel element has been calculated to be less than the stress limits for iodine stress corro-sion cracking of the fuel cladding.
Continuous monitoring of module flows and pressure drops throughout the primary system to detect the effect of predicted crud buildup have yielded results which are as expected and which remain within the assumed flow conditions in the SAR.
Since operations began, there have been four planned shutdowns for testing and maintenance. Extensive physics tests conducted at each of these times have pro-vided confirmation of the ability of the nuclear design models to calculate within required limits various parameters used in the safety analysis, e.g., bank worths, temperature and power coefficients, shudown margins, and power distributions.
For the latter, flux wire irradiations have been used. Semi-annual control drive mechanism testing has confimed the scram times used in the safety analysis along with other parameters which characterize satisfactory mechanism perfonnance.
As indicated previous.ly to NRC (NR memorandum Z#769 dated May 22, 1979, to H. Denton) one unexpected test result has been the occurrence of a larger than expected flow coefficient of reactivity during the third shutdown at 10,771 EFPH.
In this test, flow is reduced by shutting off one of the four main coolant pumps and the change l
in core reactivity is measured. At beginning of life, the reactivity change was I
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.004% R, at 7,028 EFPH (the second shutdown) the change was +.006%d t, and at 10,771 EFPH the change was +.065t4(. A program of deliberate diagnostic 2
testing was conducted to better understana the flow coefficient.
The testing demonstrated that the flow effect is distributed throughout the core, and it is likely the result of small displace:nents (0.08 in.) of tne fuel rod bundles in the radial direction probably caused by blanket module bending or translation of the blanket grids on the guide tubes resulting from hydrwlic forres. The increased value of the coefficient is jud ed td L.
dte.
t.. t..
..ill rel.u ition o: the suspension system r e.. ! !., i r..... t i..
ecy ec.,
ona i.i en.uc cyc ie; that hdve accumulated to date or fiv..
u.. v..at vi : eflecLvi f aloimet contact.
The test data i
show that the flow reactivity effect is proportional to the square of the core flow, which is proportional to force.
This is consistent with the tuodule bending and grid translation mechanisms which are self-limiting and not expected to grow significantly with time.
The protection analysis has been reassessed in light of the revised (measured) value of the flow coefficient. For the design studies, a value of three times the large:;t measured flow coefficient (.065% o f )
~~
i i
was used.
The accidents presented in the Safety Analysis Report were re-analyzed using this design flow coefficient.
As a result of these studies, it has been concluded that the thermal criteria defined in the Safety Analysis Report will be met for the worst case assumed flow coefficient where operation of all four flywheel generators is a requirement for high power operation and power / flow protection setpoints have been revised.
To monitor core behavior, a periodic test program was developed to
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periodically measure the flow coefficient to assure it is within the range assumed in the protection analysis. Periodic tests show the flow coefficient to be gradually reducing in magnitude.
III.
Factors Controlling Core Life and the basis for Extending the Core Life to Beyond 18,000 EFPH In projecting an extended life for the LWBR core at Shippingport, three interrelated areas must be considered Nuclear, Fuel Element and Thermal performance. The ultimate lifetime of LWBR is detennined by nuclear and fuel element design constraints together with retaining the primary objective of demonstrating breeding. The fuel element design objective is to prevent fuel element failures. The LWBR Safety Analysis Report demonstrated that off site exposure limits of 10CFR50, Appendix I, would continue to be met with reactor coolant fission product activity levels corresponding to the activity of 1/4% failed fuel rods. The fission l
- roduct inventory of the controlling isotopes will be less than previously calculated due to the planned reductions in power level.
If fuel element failures are detected, continued operation will be evaluated, as described in Chapter 11 of the SAR.
A.
Nuclear Performance - There must be sufficient reactivity in the core to achieve desired power levels with their corresponding equilibrium. poison (xer:en and protactinium) concentrations.
'(
Reactivity can be added to the LWBR core by: raising the bank of 12 l
movable seed modules which are the primary means of reactivity 3
contral. reducing the core power level, and reducing the average inojerator t emperature.
The 12 module bank has a maxinum withdrawal level of 8'"
above the shutdown position.
Alignment with the fixed i
blanket mdules is at 60".
The f act that the core is slightly more '
react ive th.m nmi n al (see Figure 1) and the actual o (which has led to less than equilibrium protactinium)perating history has resulted in the bank being lower than the nominal design position at the present time in life.
The vperating plan for beyond 18,000 EFPH includes recucea power and operating temperatures (see Table 1),
required by f uel element considerations, this results in the extension of the reactivity lifetime of the core beyond 18,000 EFPH.
The same nuclear design model used to predict core performance to date was used to extend the predictions beyond 16,000 EFPH.
Confidence in this estimate is based on our ability to calculate core reactivity under both full power and zero power conditions.
As for breeding perfomance, ncrainal estimates of the fissile inventory ratio were made for actual power operation to date and projected future pcwer reductions (see Figure 2).
The fissile inventory ratio (FIR), the ratio of fissile inventory to initial fissile inventory, is a measure of breeding potential. Reduced power operation beyond 18,000 EFPH will not significantly reduce the FIR from its peak value as shown by Figure 2.
Reduced power reduces losses due to neutron absorption in protactinium and xenon.
B.
Fuel Elemant and Thennal Performance - Assessments of fuel element I
and support gric capability for extended life operation indicate satisfactory performance following the planned operating program which includes specified reductions in system pressure, maximum power, and coolant average temperature (Table 1.).
These assessments are based on analysis models which have been compared to dcta from irradiation testing of over 250 LWBR type fuel rods.
Operation beyond 18,000 EFPH requires an extrapolation of irradiation test experience using analytical models.
Qualification of the thermal design procedure for LWSR was performed using over 650 critical heat flux (CHF) data points from 19 rod bundles encompassing the following range of parameters: system pressure from 2000 to 400 psia (350,108, and 231 data points at system pressures of 2000,1600 and < 1200 psia respectively), mass velocity from 0.1 to 4.0 million 1bs/hr-ft, inlet temperature from 200 to 600*F, average rod-to-rod spacings from 0.015 to 0.09 inch, (with local touching in some specific tests) and various axial and radial heat-flux distributions. Design allowances have been included such that all of the applicable CHF data are conservatively predicted.
In addition, allowances are being included to cover effects such as rods touching and increased shrinkage, wear and ovality consjdering the extended operating period and conditions.
Furthemore, the Si, power margin employed for operations up to l
l 18,000 EFPH to cover potentially unknown effects will be maintained for extended operations.
I
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4
Having discussed the qualification of fuel element and themal design procedures, specific comments concerning each period of extended life are addressed below.
During operation betaeen 18,000 and 21,000 EFPH, the limiting perfomance con-cerns are blanket rod cladding deformation over unsupported gaps, cladding stress corrosion cracking, and claddinc temperature and thermal power capability for a condition of close proximity of fuel rods.
Reduction of system pressure from 1815 to 1615 psia and of coolant average temperature from 531 to 521 F results in a large reduction of the rate of blanket rod cladding defomation, thus permitting a major increase in lifetime capability before a limiting deformation is reached. Pressure and temperature reduction also results in much lower calculated cladding temperature for rods in contact or close proximity, such that corrosion rates and cladding mechanical strength are well within limits.
Reduction in maximum power to 80% of full power capability is required to meet ther al performance limits for CHF at the reduced system pressure and to insure that pellet clad interaction (PCI) will not cause failures.
Between 21,000 and 2t,000 EFPH, the most limiting fuel concern is proximity to cladding stress limits considering both stress corrosion cracking and postulated overpower accidents. Reduction in ma
- mum allowable power from 80 to 70%'of full power reduces the calculated stress during load in-creases and for postulated accidents, and also raduces cladding temperature resulting in an incrase in cladding stress limits, :och that satisfactory performance is predicted.
I For operation beyond 24,000 EFPH, an additional factor which may be limiting is fretting wear of the fuel rod cladding. The grid spring forces are sub-stantially reduced from initial values during extended lifetime operation due to stress relaxation of the grid spring and rod diameter reduction (cladding creepdown). Results of flow testing for prototypical seed and blanket rod bundles indicate that at low spring forces (below about two pounds for blanket rods and about one pound for seed rods) cladding wear is accelerated,
)
at the lower grid level (free end) for top mounted rods. To minimize the potential for excessive wear causing breach of cladding integrity, reduction in primary coolant flow velocity is being evaluated, since reduction in flow velocity is expected to result in greatly reduced wear rate. Maximum power is reduced to 50-60% of full power capability to meet PCI and themal per-fomance limits for CHF at the reduced pressure and flow velocity.
Based on the above, the LWBR core is operating as designed and satisfactory operation to 24,000 EFPH is predicted on a conservative basis. Operation to 30,000 EFPH or more is being analyzed.
L 5
IV.
Plant _Consid_eratic.1s for LWSR Extended Life Operation A.
Pl ant Perfonnance The evaluation of the adequacy of the Shippingport reactor plant c om..ons i c. cat ed i n t i.c ! WM af ety A..alysis Report (SAR) has ba n eg en R.
tu cour to. proposea iWi'R core operating period.
This ev:botion demonstrates that the design number of heatup and s
coolco n cyt.les wili not be approachec tor the cunponents which were utilized for the previous PWR core operations and are also being used for the LWSR core.
The new components which were installed as part of the reactor coolant pressure boundary for LWBR include the reactor vessel closure head, support flange and control drive me:hanisms.
These new LWBR components were all designed to be satisfactory for the planned LWBR core operating period.
i The reactor vessel brittle fracture analysis presented in the SAR was based on 21,000 EFPH of operation for the LWBR core.
The analysis was extended to 24,000 EFPH and shows that an increase in the reference transition temperature (RTT) of only 3*F will occur.
(For 30,000 EFPH an RTT increase of only 9'F is predicted.) The pressure-temperature limits for heatup and cooldown will be revised to reflect the change in RTT from the current value cf 420*F to 429*F prior to exceeding 21,000 EFPH for the LWSR core.
The small increase in the reactor vessel RTT will not appreciably change the crack initiation and arrest depths calculated for postulated LWBR core accident conditions, and the existing casualty procedures are I
adaquate to assure the prevention of brittle fracture during accident conditions throughout the planned extended lifetime.
~
The analyses of protectable accidents presented in the SAR for operation to 18,000 EFPH have been extended for the planned LWBR operations to 24,000 EFPH and are continuing for operations to 30,000 EFPH.
These analyses are based on the revised operating conditions described in Table I.
The analyses were performed using
)
a core model which reflects the as-built core data and conservatively accounts for the observed effect of flow on core reactivity.
These analyses demonstrate that the core will continue to be protected for all postulated accidents for operation from 18,000 to 24,000 EFPH without exceeding any core thermal or fuel element limits. These results are also expected for operations to 30.000 EFPH. Appropriate changes will be required for the reactor protection setpoints corresponding to the planned changes in the operating parameters.
For the loss of coolant accident, evaluations have been perfomed for I
operations beyond 18,000 EFPH. The end of life loss of coolant accident (LOCA) analysis presented in the SAR was based on the conservative assumption that the core was underreactive such that the movable fuel assemblies would be fully withdrawn at 18,000 EFPH.
However, based on the LWBR core operating experience and the plans for LWBR core operation at reduced power levels beyond 18,000
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EFPH, it is expected that the movable fuel assemblies will not be fully withdrawn until approximately 30,000 EFPH. Thus, the heat i
6
fluxes used for the LOCA analysis and the resulting calculated cladding temperatures which are presented in the SAR are actually re,. men it i se of
- .c values which would he ex,jected for f ull power i
LW3D core operation at 30,000 EFPH.
Because of the lower movable c
'., n e, and tha reJuu i power operation, the peak
+.!
- r. e.w' ',
heat fluxes and calculated cladding temperatures for a LOCA will
,n.t ui'/
t..
.,.. e e r
.,a f"' te 3 2. 800 $ r..! t han the values 1
previously reported in the :aAR f or 10,0u0 EI PH.
In addition, doses resulting f rom the postulated LOCA beyond 18,000 EFPH do not exceed 10CFR100 limits since the fission product inventory of the controlling isotopes will be less due to the planned reductions in power level.
For operation beyond 18,000 EFPH, the fuel rod internal pressures will increase due to increased fission gas release.
However, for operation from 18,000 to 30,000 EFPH, the net effect of the lower cladding temperatures and the increased fuel rod internal pressures is that the total fuel rod diameter increase and thus the amount of channel closure during the worst case LOCA will be less than previously reported in the SAR.
Thus, for operation beyond 18,000 EFPH, the consequences of a LOCA will be less severe than the results )reviously reported in the SAR at 18,000 EFPH, and the requirements in Appendix K to 10CFR50 will continue tu be satisfied.
B.
Planned Testing, Inspections and Maintenance The LWPR Test Program, which is described in the SAR, includes the
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tests which were performed prior to initial criticality and also the tests which are to v' e perfomed periodically throughout LWBR operations.
The testing requirements and guidelines in 10CFR50, applicable Regulatory Guides, IEEE Standards, and the ASME Code were utilized in the development of the LWBR Test Program.
It is planned to continue the periodic test program during the extended operations,
beyonc 18,000 EFPH.
It is noted that an overall container leak rate test was perfonned for the Shippingport containment prior to LWBR core operation. The need to perfom another overall container leak rate test to support the planned extension in LWBR core operations beyond 24,000 EFPH is currently being evaluated.
An inservice inspection program was perfomed for the Shippingport Plant prior to '.WBR core operations. As noted in the SAR, the requirements of Section XI of the 1974 edition of the ASME Code were utilized as a guide in selecting items fcr the inservice inspection program. There were no indications found which required repair as a result of the Shippingport inservice inspection program.
To support an extension in LWBR core operations beyond 24,000 EFPH, an additional inservice inspection for the Shippingport Plant is being evaluat ed.
The inservice inspection would include volumetric
, inspection of selected pipe welds, as well as specific visual and hydrostatic tests. The pipe welds selected for volumetric inspection would include a portion of the welds inspected prior to initial LWBR core criticality to verify that there is no degradation of these welds. The pipe welds selected would include k
representative welds of various sizes in the reactor coolant system and connecting, system piping, f
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The program for the performance of required and preventive maintenance which has been utilized for the Shippingport Plant will be continued for the extended LWSR core operations.
The maintenance program is designed to pro-vide for safe and reliable plant operation by performing preventive maintenance on a timely, periodic schedule.
As have been documented in annual environmental monitoring reports, releases of radioactivity from the Shippingport Atomic Power Station continue to be far below applicable standards with the result that the LWBR core operation has not adversely affected the surrounding environment. This result is not expected to change for future operations and the conclusions contained in the LWBR Program Environmental Statement ERDA 1541 concerning environmental aspects of LWSR core operation in the Shippingport Atomic Power Station remain valid.
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8
TABLE 1 LWBR OPERATING PLAN (l)
Lifetime Intervals 0-12,000-18,000-21,000-24,000- (4) 12,000 EFPH 18,030 EFPH 21,000 EFPH 24,000 EFPH 30,000 EFPH 1.
Power (2)
% of 236 MWth 100 100 80 70 50-60 1815 1615 1615 1615 2.
Pressure-psia 2000- (3) 1815 3.
Average Temperature
'F 531 531 521 521 521 i
(1) Actual times in life when planned reductions will occur will be specified to be consistent with planned semi-annual shutdowns for testing and maintenance.
(2) 100% power corresponds to 236.6 MW(th) and 62 net MW.
(3) NR memoranda Zi681 dated May 11, 1978 and Zi769 dated May 22, 1979, informed NRC that pressure would be reduced from 2,000 to 1,815 psi to optimize per-formance of blanket rod cladding.
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(4) Operating parameters for operations beyond 24,000 EFPH are currently being evaluated.
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