ML20052F340
| ML20052F340 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/07/1982 |
| From: | Ronaldo Jenkins CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20052F337 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8205120359 | |
| Download: ML20052F340 (13) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
) Docket No. 50-155-OLA CONSUMERS POWER COMPANY
) (Spent Fuel Pool
)
Modification)
(Big Rock Point Nuclear Power Plant)
)
4 AFFIDAVIT OF ROLFE B. JENKINS ON REACTOR VESSEL EMBRITTLEMENT I,
Rolfe B. Jenkins, of lawful age, being first duly sworn, do state as follows:
1.
I am employed by Consumers Power Company as a Senior Staff Engineer in the Operating Services Department in Jackson, Michigan.
I have a Bachelor of Science Degree in Mathematics and a Master of Science and Doctor of Philosophy Degree in Engineering Mechanics, all from Michigan State University.
I am a registered professional engineer in the State of Michigan.
2.
Upon being graduated in 1973, I worked three years at Westinghouse Electric Corporation - Bettis Atomic Power Laboratory in West Mifflin, Pennsylvania.
While working in the Light Water Breeder Leactor project, I was responsible for core grid design, core component manufacturing evaluation, core component assembly, and specific structural analysis for the evaluation of the as-built core structural components.
820512O359 820500 PDR ADOCK 05000155 C
3.
I joined the Operating Services Department of Consumers Power Company in 1976.
As part of my duties, I have been responsible for the implementation of reactor vessel surveillance programs for both of Consumers Power Company's operating nuclear plants in accordance with 10 C.F.R. Part 50, Appendix H.
I am also responsible for assembling the data resulting from those programs so as to develop pressure-temperature limits for the reactor vessels in accordance with 10 C.F.R.,
Part 50, Appendix G.
In order to maintain an awareness of new data and technologies associated with the radiation damage-fracture toughness issues, I am affiliated with the Electric Power Research Institute Nuclear Division Structural Mechanics Subcommittee of the Systems and Materials Task Force.
In addition, I interact with the Metal Properties Council Subcommittee 6 on nuclear materials.
4.
The purpose of this affidavit is to respond to Atomic Safety and Licensing Board questions concerning reactor vessel embrittlement in its memorandum dated March 25, 1982.
The questions are:
(1)
Whether increases in the pressure-temperature li aits of the reactor vessel.nay affect the safety or operability of the entire reactor system; and (2)
Whether the reactor pressure vessel will
. continue to be safe, for its expected life, with respect to thermal shock which might occur if the emergency core cooling system or other makeup water systems had to be activated in response to a loss-of-coolant accident.
5.
The questions listed in paragraph 4 are follow-up questions to the following Licensing Board question:
We would like to know whether it is necessary for us, in considering either environmental or safety issues, to know whether there is a substantial change in the Staff's view of the life expectancy of the reactor vessel or the life expectancy of the Big Rock Point Plant as a result of the embrittlement problem.
(Tr.
290).
Although this question was directly posed to the NRC Staff, the Board indicated that it was also interested in the Licensee's view on this question.
(Tr. 290).
The question t
was addressed by the Affidavit of Mr. Emch of the NRC Staff dated March 22, 1981.
I agree with his conclusion that reactor vessel embrittlement will not shorten the life of Big l
Rock Point Plant.
6.
Question 1:
Whether increases in the pressure-temperature limits of the reactor vessel may affect I
. the safety or operability of the entire reactor system.
Response
It has long been understood that low-alloy steels such as those from which reactor vessels are fabricated realize certa!.n deleterious metallurgical effects when sub-jected to a high-energy neutron environment.
The welds and heat-affected zones from the welding process also are subject to the same metallurgical effects.
High energy neutrons have the effect of reducing the ductility of the vessel steel and welds over a period of time.
The loss in ductility is reflec-ted in a lesser capacity of the steel or welds to absorb energy.
The energy absorption capacity is desirable so as to ensure that any flaw which might exist in a vessel can, at worst, propagate in a stable manner and result in a leak rather than in a catastrophic brittle failure.
In order to monitor and evaluate the effects of neutron irradiation, 10 C.F.R. Part 50, Appendix H, requires that every nuclear power plant facility being licensed today have a reactor vessel surveillance program meeting the require-ments of that Appendix.
Although surveillance programs can take many forms, all surveillance programs utilize a number of surveillance capsules which in themselves contain neutron exposure monitoring devices.
They also utilize plate, weld,
9 and heat-affected zone weld material specimens which can be tested at some future date to evaluate the damage done to the vessel itself as a function of neutron exposure.
These capsules are withdrawn from the inside of the vessel periodi-cally on a schedule consistent with the requirements of 10 C.F.R. Part 50, Appendix H.
The data from surveillance capsule testing provides a basis for projecting neutron exposure and vessel material damage as a function of time.
These data and projections are then formatted into reactor coolant system pressure-temperature limits per the ASME Boiler and Pressure Vessel Code
(" ASME Code"), Appendix G, as invoked by 10 C.F.R. Part 50, Appendix G.
The pressure-temperature limits are typically drafted as operating curves by the utility personnel and proposed to the NRC Staff as Technical Specification changes.
When the Staff's concurrence is obtained, the pressure-temperature limits are implemented.
As has been stated, the reactor surveillance program and the resulting reactor vessel pressure temperature limits exist to ensure that the vessel materials are always in a ductile rather than a brittle condition.
For a given material, neutron exposure tends to reduce ductility at a given temperature.
Thus, reactor vessel pressure-temperature limits seek to increase the, metal temperature at a given I
pressure to compensate for neutron exposure so that the l
material ductility will not be compromised during the vessel operating life.
The Big Rock Point Plant has continued over the years to implement a reactor vessel surveillance program which meats the requirements of 10 C.F.R. Part 50, Appendix H.
To date, five surveillance capsules of irradiated specimens and a set of unirradiated material test specimens have been evalua-ted.
Neutron dosimetry data from the five irradiated capsules have been analyzed.
The neutron dosimetry and material damage data are,- therefore, well defined for the Big Rock Point reactor vessel and apply to neutron fluence levels extending well beyond the end of design life.
About-two years ago, Big Rock Point removed the first capsule that it had removed in approximately ten years.
The capsule was tested by Westing-house Electric and the test report was transmitted to the NRC Staff along with the Technical Specification change.
The testing showed that the rate of neutron damage in the limiting reactor vessel welds had slowed considerably from that deter-i mined from early-in-life exposure.
It has been hypothesized that this very beneficial result *is due to the low nickel content in the welds.
Appendix G of the ASME Code was developed in between the time of removal of the fourth and fifth surveillance capsule.
Because revisions to pressure-temperature limits seldom are drafted without an associated l
l l
i l
l
. capsule removal and test, the Technical Specifications for pressure-temperature recently drafted are the first such set formatted to ASME Code, Appendix G guidelines.
Because of the different formatting, it is very difficult to compare former and revised pressure-temperature limits.
The recently revised pressure-temperature limits written by Consumers Power Company incorporate an l
evaluation of the recently removed surveillance capsule.
The revision increases the minimum reactor coolant temperature allowed at any given pressure.
These limits employ linear elastic fracture mechanics with test data from the reactor vessel surveillance capsules.
The heatup and cooldown limits are also based upon the assumption of very slow temperature transients, a hypothetical one-quarter wall thickness initial flaw on either the inside or outside vessel wall and a factor of safety of two with regard to pressure stress.
For a maximum heatup or cooldown rate of 100'F/hr., the pressure-temperature limits require a minimum reactor coolant system i
l temperature of approximately 340'F at the design pressure of l
1700 psi when the core is critical.
The reactor coolant system typically operates at 1350 psi and approximately 580*F.
All reactor coolant system cv.aponents, including pipes, valves, steam drum, and the reactor vessel itself, are de-signed to a pressure-temperature rating of 1700 psi and 650'F.
i l
l
, Any steel component designed for a given pressure-temperature rating can sustain significantly higher pressures at a lower temperature (340*F as compared with 650*F).
In addition, surveillance capsules attached to the thermal shield of the reactor have been subjected to more neutron exposure than the vessel wall will be at the end of its operating life.
The testing of these capsules has proved that future shifts in pressure-temperature limits will in no manner compromise the capacity of the reactor coolant system to operate within the design limita of any of its components.
The revision of the reactor vessel pressure-temperature limits has not changed either the maximum pressure or the maximum temperature at which any component of the reactor coolant system is allowed to operate.
The pressure-temperature limits concern themselves only with minimum temperatures at existing allowable pressures.
These tempera-tures are well below normal operating temperatures and even farther below the design value of 650*F.
Therefore, the safety of the reactor coolant system is not compromised by the revision to the pressure-temperature limits.
The revision to the pressure-temperature limits does have an effect on plant operability.
For any plant, a revi-1 s, ion requires a higher temperature at a given pressure for heatup, cooldown, and hydrotest conditions.
Therefore, it i
1 i
- require significantly more hours for the plant to achieve hydrotest pressure and to go into operation.
Question 2 Whether the reactor pressure vessel will continue to be safe, for its expected life, with respect to thermal shock which might occur if the emergency core cooling system or other makeup water systems had to be activated in response to a loss-of-coolant accident.
Response
The reactor vessill pressure-temperature limits are based on slow thermal transient conditions.
They are not based upon a pressurized thermal shock scenario.
However, the data and methodology employed in the development of these limits provide insight into the reasons why pressurized thermal shock is not an issue at Big Rock Point.
An under-standing of this issue can best be achieved by comparing the Big Rock Point reactor vessel and operating system with that of a typical PWR.
A comparison with a typical PWR is made because thermal shock is generally perceived to be only a PWR concern.
More specifically, such a comparison must consider the forcing functions which produce loads in the vessel (pressure and thermal wall gradients) and the resistance of the vessel to these loads (material quality-toughness or ductility).
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ A.
Reactor Vessel Pressure Stress The Big Rock Point Plant normal operating pressuring stress is 60-65 per cent that of a typical PWR.
Since pressure stress warrants a safety factor of two (per ASME Code, Appendix G) in the development of pressure limits with respect to normal operating cycles, its significance in the propagation of any flaw which might exist is clear.
B.
Reactor Vessel Thickness The reactor vessel wall thickness at Bic Rock Point is about two-thirds that of a typical PWR (5.625 inches).
The thinner wall implies that the thermal gradient across the vessel wall due to a given cold water injection would be significantly lower than in a PWR.
Therefore, the tensile stress at the vessel wall inner diameter would be perhaps 50 per cent less than that in a typical PWR under a similar cold water injection condition.
C.
Reactor Vessel Material The Big Rock Point Plant reactor vessel base metal, weld metal, and weld
heat-affected zone materials are of very good initial quality, which is charac-teristic of reactor vessels fabricated by i
Combination Engineering.
This metal is less susceptible to radiation damage than USNRC Regulatory Guide 1.99 trend curves, which have been provided by the Staff for facilities which have limited operating time and, therefore, not enough surveil-lance capsule data to provide their own t' rend curves.
This has been proved by testing of the five surveillance capsules and the unirradiated test samples.
This implies that the vessel wall material will be more capable of arresting (rather than propagating) any initiated flaw than materials in PWR's whose properties form much of the basis for USNRC Regulatory Guide trend curves.
D.
High Pressure Safety Injection Big Rock Point does not have a high pressure safety injuction system.
The Big Rock Core Spray System provides safety injection at a pressure of less than 150 l
l
. psi, which is well below the operating pressure of 1350 psi.
Therefore, the coincident occurrence of tensile stresses on the reactor vessel inner diameter due to temperature gradients and operating pressure effects does not occur during a thermal shock due to cold water injection.
This is perhaps the most significant of the four considerations discussed here.
Vulnerability to thermal shock is based on the simultaneous occurrence of high pressure stress and high thermal gradients being imposed upon a relatively nonductile material.
The Big Rock Point reactor vessel experiences i
relatively low pressure and thermal gradient stress which would not occur simultaneously in the event of a cold water injection.
In addition, the loads which could develop would be imposed on very high quality vessel material.
Because of i
such considerations, thermal shock is not judged to be a safety issue in BWR's in general and Big Rock Point Plant in i
l particular, and the margins of safety associated with 10 C.F.R. Part 50, Appendix G, pressure-temperature limits are not compromised.
Therefore, it is concluded that the reactor
~ _..
vessel is safe from the effects of thermal shock as might be imposed by a cold water injection in response to a loss-of-coolant accident.
M$
V i
i Subscribed and sworn to before me this 74 day of May, 1982.
afa m
.273 Notary ublic My e
-- Empise Jamesy I.1987 1
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