ML19325E015
| ML19325E015 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/19/1989 |
| From: | Lorion J CENTER FOR NUCLEAR RESPONSIBILITY, LORION, J. |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML19325E014 | List: |
| References | |
| OLA-4, NUDOCS 8910310197 | |
| Download: ML19325E015 (148) | |
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1 UNITED STATES OF AMERICA 69 0% 23 P4 '33 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARDr-DO i
L In the Matter of
)
)
Dockets Nos. 50-250 OLA-4
' FLORIDA POWER & LIGHT COMPANY
)
50-251 OLA-4
)
F (Turkey Point Plant,
)
(Pressure / Temperature Limits)
Units 3 and 4)
)
t.
)
i INTERVENORS' RESPONSE TO LICENSEE'S MOTION FOR
SUMMARY
DISPOSITION OF INTERVENORS' CONTENTIONS Pursuant to 10 C.F.R. 2.749, Intervenors, the Center for Nuclear Responsibility and Joette Lorion (Intervenors), hereby file their response to Licensee's motion for summary disposition in the above captioned proceeding.
In support of this
- response, Intervenors have attached "Intervenors Statement of Material Facts As To Which There Is A Genuine Issue To Be Heard With Respect To Intervenors' Contentions" and the letter of Dr.
George Sih on Contention 2 dated October 18, 1989 (Sih Letter, Attachment A).
As discussed
- below, the Intervenors contend that there is a genuine issue of material fact regarding the matters set forth in the attached statement and affidavit, and that the Licensee is not L
entitled to a decision in its favor as a matter of law and summary judgment should be denied.
8910310197 091019 PDR ADOCK 05000250 0
PDR L
I.
BACKGROUND OF THIS PROCEEDING On October 19,
- 1988, a
notice was published in the Federal
)
Register announcing' the proposed issuance of amendments to the 4
Technical Specifications for Turkey Point Units 3 and 4.
53 Fed.
Reg.
40988.
The Proposed amendments would modify the
. pressure / temperature limits for the reactor coolant system and the pressurizer for each unit.
On Novemeber 17, 1988, the Center for Nuclear Responsibility, Inc.
(" Center")
and Joette Lorion, collectively referred to herein as "Intervenors",
filed with the Nuclear Regulatory Commission
("NRC")
a Request for Hearing and Petition for Leave to Intervene
-(" Petition")
concerning the Florida Power & Light ("FPL") amendment
. request.
On January 10, 1989, the NRC Staff issued Amendment Nos. 134 and-128 to the operat1ng licenses for Turkey Point, Units 3 ano 4 respectively, revising the pressure / temperature ("P/T") limits for the. Turkey Point units along with their Safety Evaluation and Final Determination of No Significant Hazards Consideration.
The Intervenors then submitted their
" Amended Request for Hearing and Petition for Leave to Intervene" on February 17, 1989, 1
which listed three Contentions that Intervenors asked to be admitted i
L l
for litigation in this proceeding.
On March 21, 1989, the Atomic L
Safety and Licensing Board (Board) held oral argument on the l
L contentions.
Subsequently, on June 8,
1989, the Board issued an l.
l Order which denied Contention 1 and accepted portions of Contentions 1
i l
l l l
1
2 and 3.
On September 8,
- 1989, after a
meeting with the Licensee, i
Intervenors withdrew Contention 3 from this proceeding.
Finally, on September 11,
- 1989, the Licensee filed their Motion for Summary Disposition of Intervenor's Contentions.
II. LEGAL STANDARD FOR
SUMMARY
DISPOSITION The summary disposition procedure should be utilized on issues where tiiere is no genuine issue of material fact to be heard so that evidentiary hearing time is not wasted on such issue 6.
Statement of Policy on Conduct of Licensing Proceeoings, CLI-81-8, 13 NRC 452, 457 (1981); 11sconsin Electric Power Co. (Point Beach Nuclear Plant, Unit 1),
ALAB-696, 16 NRC 1245, 1263 (1982); Houstor Lichtino and Power Co.
(Allens Creek Nuclear Generating Station.
Unit 1 ),
ALAB-590, 11 NRC 542, 550 (1980).
It is the movant, not the opposing party, which has the burden of showing the absence of a genuine issue as to any material fact.
Cleveland Electric 111uminatino Co.
(Perry Nuclear Power Plant, Units 1
and 2), ALAB-443, 6 NRC 741, 753 (1977).
Since the moving party has the burden to show initially the absence of a genuine issue concerning any material fact, where the evidentiary matter in support of the motion does not establish the absence of a genuine
- issue, summary judgment must be denied even if no opposing l
l evidentiary matter is presented.
edickes v.
Kre}s & Co., 398 U.S.
- 144, 160 (1970).
However, if the motion for summary disposition is properly supported, the oppsition may not rest upon
" mere l L k
I l
r 1
I I
i
[
allegations or denials"; rather, the answer "must set forth specific facts showing thst, there is a genuine issue of fact."
Vircinig
)
and 2),
j l
Electric and Power Co. (North Anna Power Station, Units 1
'ALAB-564, 11 NRC 451, 453 (1980),
i l
A.
BACKGROUND
)
There is a
- high, increasing likelihood that someday
)
i
- soon, during a seemingly minor malfunction at any of a dozen or more nuclear power plants around the United States, the steel vessel that houses the radioactive core is going to crack like a
piece of glass.
The result will be a core i
meltdown, the most serious kind of nuclear accident.
1 Demetrios Basdekas, NRC Safety Engineer "The Risk of a Meltdown,"
New York Times (March 29, 1982), (Exhibit 1).
2 Two facts have been known since our ne' ion undertook the j
commercial development of nuclear power. 1) the absolute integrity I
of the large steel vessel tht houses the core and contains the cooling water for the reactor is central to protecting the health ar.d safety of the adjoining community and the environment and 2) all
- metals, including steel, become embrittled overtime as a result of continued exposure of neutron irradiation. (Exhibit 2).
Nuclear plant pressure vessels are fabricated from ferritic steels.
At Turkey Point, for instance, large sections of eight inch thick steel are welded together circumferentially to form the i
I housi'ng for the reactor core.
The safety of the public depends on the ability of the e
m
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1 1
i materials in the vessel and the welds to maintain their fracture b
toughness.
Fracture toughness is a material property that enables
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the material to resist brittle fracture when stressed.
An adequate level of fracture toughness provides the assurance that small flaws or cracks will not propagate in a " brittle manner" as a result of stresses caused by reactor
- heatup, cooldown and/or abnormal 1
It is well known that for steels used in nuclear reactor pressure vessels and their
- welds, three considerations are importent.
- First, fracture toughness increases with increasing temperatures;
- second, fracture toughness decreases with increasing load
- rates, and
- third, fracture toughness decreases with neutron irradiation.
In recognition of these considerations, power reactors are operated within restriction imposed by the Technical Specifications on 4he pressure during heatup and cooldown operations.
These restrictions assure that the reactor vessel will not be subjected to i
that combination of pressure and temperature that could cause l
brittle fracture of the vessel if there were significant flaws in the vessel material. The effect of neutron radiation on the fracture toughness of the vessel material is accounted for in developing and revising these Technical Specification limitations over the life of the plant.
The pressure / temperature liniits, which are the subject of this proceeding are just such restrictions.
Additionally, there is another issue to consider where fracture I
t i
[ f
r.
l' t
toughness is concerned.
That is the fact that in many of the older nuclear
- plants, such as Turkey. Point, high levels of copper and nickel were used to fabricate the welds of the vessels and in some cases the vessels themselves.
These elements were later shown to result in greater irradiation damage to the vessel material than had been initially expected.
Irradiation damage in these plants caused a
shift in the fracture toughness curve to higher temperatures, and therefore, increased the possibility of a nonductile failure.
This is so because as metal embrittles, it loses the property of " ductility" and must be kept at increasingly high temperatures in order to retain adequate ductility to avoid cracking or shattering in response to stresses or shocks. (Exhibit 2).
In
- 1981, the Nuclear Regulatory Commission became concerned about the extent of the embrittlement problem at some of the nation's older nuclear power reactors. This concern was manifestec as a
result of NRC Safety Engineer, Demetrios Basdekas' warnings that some of the more embrittled reactors with high copper contents in their welds could shatter from pressurized thermal shock (PTS) and endanger the commun1 ties in which the plants were locateo.
(Exhibits 1 and 3). As part of the NRC's investigation of the (PTS) phenomenon, they sent letters pursuant to 10 C.F.R. 50.54 to Licensees whose fracture toughness of their reactor pressure vessels were approaching levels of concern. (Exhibit 4 ).
Florida Power and Light Company received just such a letter concerning the Turkey Point Unit 4 reactor. FPL was asked to submit '
i
I I
plant specific information to the NRC in 150 days in lieu of i
licensing action.
(1d2).
The Licensee was not asked to submit information on Unit 3, nor did the NRC sing'le out Unit 3 as one of the nuclear power reactors that concerned them.
{
- Yet, when the Licensee responded to the NRC's 50.54 letter on August 23,
- 1981, concerning a Quest 1on the NRC had proposed as to the reference temperature nil-ductility transfer value (RTNDT) for Unit 4,
the Licensee responded that the value they had provided the NRC was based on Unit C
data which had been shown to be more I
representative of Unit 4 than the surveillance capsule that had been removed from Unit 4 (Exhibit 5).
l The surveillance capsules that the Licensee was refering to
{
were samples of weld material that they and other licensees are I
required to install in each reactor vessel so that they can be periodically withdrawn and tected to determine the actual extent of the embrittlement that has occurred in the specific reactor vessel.
(Exhibit 2).
These samples are recuired by 10 C.F.R. Appendices G and H to be withdrawn periodically and subjected to a
process known as "Charpy" tests.
In these tests, specimens are heated to different temperatures and then struck to determine the temperature at which the metal shatters or cracks in order to drctermine the extent of embrittlement and the minimum temperature that must be maintained in order to assure the metal retains sufficient ductility to resist anticipated shocks (19.). The danger-point occurs at the temperature. -.
^
at which the metal loses its ductility (or arrives at
n i l t
ductility"). The Commission and the industry use the term " reference temperature for nil ductility transit. ion " abbreviated as "RTwot',
to identify this danger-point.
In 1974 and 1975, the f.icensee removed weld mctal capsules T from Turkey Point Units 3
and 4 and Charpy tests were performe separately on the samples from each unit. (Exhibit 6 and 7).
The central document necessary to demonstrate the basis for Intervenors continuing concerns regarding Unit 4
is a
report submitted by the Southwest Research Institute (the " Institute")
entitled Pressure Temoerature Limitation for the Turkey Point Unit Nos.
3 and 4
Nucigpr Power Plants, SWRI Project No, 02-4363-039 (June 30, 1976).
(Exhibit 8).
The Institute had conducted Charpy i
?
tests on metals containeo in a
capsule taken from Unit No. 4 (Exhibit 7). Materials contained in the capsule taker, from Unit No.
3 had been tested by the Westinghouse Electric Corporation (Exhibit 6).
Thereafter, the Institute was asked to project the separate "heatup and cooldown limit curves
- for the vessels for Units No. 3 and No.
4 applying the Commission's prescribed computational criteria to the separate test results on materials taken from each of the two units (id.). The Institute's summary of its results, set forth in the
- margin, illustrates the dramatic difference in embrittlement found in the Unit No. 3 samples from that found in tne Unit No.
4 samples after less than three years operation (Exhibit 8).
The data also suggests that, as early as 1976, the Commission 4
l
\\
i and FPL were aware that the best available data indicated that the embrittlement occurring in Unit No.
4 would require that the l
temperature of that vessel be maintained at well-above 300 degrees F l
l to maintain accepthble ductility before the Unit had been in operation for the equivalent of ten effective full power years i
("EFPY) ( jf. ).
The values of RTwet for the beltline regions of Turkey Point Units Nos.
3 and 4
were derived from (1) the surveillance program test results.
(2) computed ratics of i
fast flux at the 1/4 and 3/4 locations in the vessel wall, and (3) trend curves in RTwot as a function of neutron fluence (E 1 HeV).
A summary of these values is as follows:
Unit Operating RTNDT RTNOT l
No.
Period at 1/4 T at 3/4 T 3
5 EFPY 194 deg.F 131 deg,F 3
10 EFPY 236 deg.F 159 deg,F 4
5 EFPY 281 deg.F 188 deg.F 4
10 EFPY 342 deg.F 230 des.F
- EFPY : Effective Full Power Year E.
Norris and J.
Unruh, Ergisure-Tamperature_ Limitations for the Turkey Point Unit Noga 3& 4 Nuclear Powef Plan _t at 27 (SWRI Project No. 02-4383-039 (June 30, 1976). (Exhibit 8).
l How does a reactor whose pressure vessel that the Institute's 1976 report projected would exceed the NRC's own 300 deg. F screening criterion after less tPan ten Efective Full Power Years 1
(EFPY) continue to operate
?
The public record suggests that continued operation is the product of legal alchemy rather than technical progress.
The legal alchemy was achieved simply and in a 1 1
I L
i-I manner that would have been impossible if the NRC Staff hac not allowed the Licensee to calculate the RTNDT for Unit 4 based upon
" Unit 3
data" in response to the Commission's 1981 60.54 letter.
(Exhibits 4 and 5).
Thus, it appears that the NRC Staff allowed the Licensee to use an integrated surveillance program to calculate the embrittlement of Unit 4
long before they confirmed the practice on April 22, 1985 when they issued a license amendment to FPL which allowed them to use an integrated surveillance program to calculate radiation damage to the Turkey Point reactor vessels.
As did the Licensee, the NRC appears to have ignored the actual differences in levels of embrittlement disclosed by the 1976 reports for Units 3 and 4, and has authori:ed FPL to continue operating Unit 4
so long as Turkey Point Unit 3
meets the Commission's embrittlement criterion.
The record before this Board now suggests that the NRC Staff continues to ignore tha fact that the only data ever derived from weld metal tests for Unit 4 demonstrates that it is non-conservative and improper to calculate the ART and revise the P/T limits for Unit 4 based primarily on data from the less severely I
affected Unit 3.
Intervenors contend that neither the Licensee nor the Staff have given the Board proper justification for thir decision not to test Unit 4's capsule V weld metal specimen in order to revise the P/T limits for that unit. A decision, which if sanctioned by this l
l
- Board, could make a rupture of the reactor pressure vessel with its --
I i
enormous public health and safety consequences more probable.
B. ISSUES RELATED TO CONTENTION 2:
Contention 2 states as follows:
That the revised temperature / pressure limits that have be6t set for Turkey Point Unit 4 are non-conservative and will cause that reactor unit to exceed the requirements of General Design Criterion 31 of Appendix A to 10 CFR Part 50, which requires that the reactor coolant pressure boundary be designed with a
sufficient margin to ensure that, when stressedd under operating, maintenance,
- testing, and postulated accident conditions, (1) the boundary behaves in a non-brittle manner and (2) the probability of a rapidly propagating fracture is minimized.
Petitioners contend that the new pressure / temperature limits could cause the reactor vessel to exceed these requirements because the Licensee has based its calculation of the predicted RTNot for Unit 4 partly on surveillance capsule V
test results from Turkey Point Unit 3 rather than prediction the RTNet for Unit 4 based on Unit 4 capsule V surveillance capsule data--a practice which is not scientific, not valid, and could cause the Unit 4 reactor to behave in a brittle manner which would make the Chances of a pressure vessel failure and resultant meltdown more likely.
l Petitioners contend that predictions of RTNOT and pressure / temperature limits derived from the shift in nil-ductility transfer should be based only on plant-specific Unit 4
- data, especially in light of the fact that the only tests ever performed on Unit 4 weld specimens demonstrated that the weld material in the Unit 4 vessel was 30t more brittle than that of Unit 3. Because Unit 4's weld material is more embrittled, Petitioners contend that the FPL Integrated Surveillance program does not meet the Requirements of 10 CFR Appendix G Parts V.A and V.B.
and 10 CRF Appendix H,
including Appendix H Parts IIC and IIIB.
- Finally, Petitioners contend that the surveillance capsule V l
for Unit 4
should be tested to establish the new pressure / temperature limits and should the testing indicate that the RTNOT for Unit 4 has passed the 300 deg F screening criterion set by the NRC, Unit 4 should be shut down unit 11 it is demonstrated that the Unit 4 reactor pressure vessel I
can maintain its integrity beyond this limit.
L l
The pressure / temperature limits for Turkey Point Units 3 and 4 ru l
)
i l
3 are among the most critical limiting concitions of operation because they define the permissable operating envelope during l
reactor heatup, cooldown, criticality, and testing and are designed to wnsure the integrity of the reactor pressure vessel, a critical piece of safety equipment, j
According to 10 C.F.R.
Appendix G, the pressure / temperature limits must.be predicted based on the results of certinent radiation effect studies that predict the effects of neutron irradiation on pressure vessel embrittlement. These limits are required to be based on the most limiting nil-ductility reference temperature (RTNDT) f or the respective reactor units, i
As explained earlier, the reference temperature is the point at j
l which the pressure vessel metal loses nearly all of its ability to withstand shock.
- Thus, it is necessary to accurately and conservatively account for the effects of irradiation and other i
l factors on the RTNDT of the pressure vessel in order to set l
i conservative P/T limits that will protect the public from a brittle j
fracture of the vessel and subsequent meltdown of the reactor core.
l l
In order to meet the requirements of Appendix G, 10 C.F.R.
Appendix H requires the Licensee to estab'.ish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel and test them to determine shifts in the RTNDT. This I
calculated shift in the fracture toughness of the vessel material l
due to neutron irradiation damage is called the Adjusted Reference Temperature or (ART).
Appendix H
also allows an integrateo j
l l l
l surveillance program for multiple reactors located at a single site j
i on an individual case basis depending on the degree of commonality and the predicted severity of irradiation.
[
l f
contention 2
primarily contends that the current pressure / temperature limits that were set for Turkey Point Unit 4 do
[
not meet the requirements of Appendices G and H, and that these limits are non-conservative and could cause the Turkey Point reactor f
Unit 4 to exceed the General Design Criterion 31 of Appendix A to 10 C.F.R. Part 50.
Intervenenors base their belief on the following issues of l
fact:
1.
The Turkov Point Unit 4 oressure/temocraturg_ limits should be set usino clant acacific data.
Intervenors have contended throughout this proceeding that the revised Turkey Point Unit 4 P/T limits should have been based on the results of plant specific surveillance capsule test data.
l Intervenors base their contention on the Pacific Northwest Laboratory Report NUREG/CR-2837 entitled PNL Technical Review of Prestyrized Therm 31 Shock
- Issuti, July 1982 which states that
" evaluating the failure probability of any nuclear pressure vessel is very complex. The evaluation must be plant-specific to allow for differences in material properties of the plant components, systems, configuaration, operating procedures, and dosimetry history."
(Exhibit 9
at 1.1)
The report also states that " predicting the i
material properties of plant-specific reactor vessels requires an _ _
P l
accurate knowledge of neutron exposures of metallurgical test i
specimens and an accurate knowledge of the neutron exposure of f
plant-specific pressure vessel components." (ig. at 5.11).
7 This view is also supported by NRC Safety Engineer, Demetrios Basdekas in a memo to Commissioners Gillinsky and Ahearne, re: Staff Report on PTS, dated December 3, 1982, wherein Basdekas states that a
meaningful PTS assessment may be performed in a plant-specific basis only.
(Exhibit 10 at p.3).
One should note that botn the analysis of P/T limits and the analysis or screening criterion for pressurized thermal shock (PTS) depend on the changes in the fracture toughness of the beltine material.
- Finally, this view is further supported by Dr. George Sih, Director of Fracture Mechanics at Lehigh University who states in a letter to Intervenor's former attorney Martin H.
Hodder, dated October 10, 19ES, that:
The rate at which the beltline weld mater 1a1 deteriorates and/or embrittles depends on the combined effects of irradiation and pressurized thermal shock. It is plant-specific in the sense that the influence differs l
inherently from one unit to another. In other words, the metallurgical properties alone cannot determine the damage behavior of the welds.
The loading history' plays a major rcle. Unless the rates of irradiation, fluctuation in thermal l
gradients and tiri.: veriation in pressure are exactly the same i
for both Units No.
3 and No.
4, one is not justified to assume that data collected in Unit 3 could Le applied to predict the behavior of Unit No. 4.
Hence, conslusions drawn on change of RTHof for Unit No. 4 based on the data of Unit l
No. 3 cannot be considered valid.
(Exhibit 11 at 2) l l
The need for plant specific data to be used to calculate the adjusted reference temperature (ART) to revise the Unit 4 P/T limits l
+
^~
1 i
I I
is especially significant in light of the fact that the only known test data concerning the actual embrittlement of Unit 4 demonstrated thac the neutron damage to the pressure vessel welds in Unit 4 was far greater than anticipated and far greater than the embrittlement l
of the reactor vessel for Unit 3.
(Exhibit 8)
- Thus, Intervenors find it incredible that the NRC Staff would allow the Licensee to use data from the less severely affected Unit 3
combined with the original Unit 4 data, (which results in a l
smearing and dilluting of the data), to predict the P/T operational limits for Unit 4 The central issue necessary to demonstrate the basis for Intervenors' continuing concerns is the Licensee's Integrated Surveillance Program. Thus, Intervenors will address the majority of their issues of fact in their discssion of that program.
l l
2.
Intervenors contend _that the Licensee never met the reaui rements of the Intenrated Surveillance Pronram._and thev still l
l don't meet the reauirements.
As explained
- earlier, since the fracture toughness of the reactor vessel changes as the vessel is exposed to neutron irradiation, it is necessary to periodically recalculate the P/T L
limits to account for changes in the fracture toughness of the l
reactor vessel.
This change is the fracture toughness, or adjusted reference l
l temperature (ART) is calculated by removing surveillance capsules of. - - -. -
weld material from the reactor units and performing charpy tests on the surveillance cpecimens. Appendices G and H of 10 C.F.R.
recuire that licensee's periodically remove and test surveillance capsules I
to determine the shift in RTNDT.
)
Appendix H
allows in some cases for the reactor surveillance programs to be combined and/or integrated, According to Appendix H,Section II.C there are certain criteria to be used in evaluating whether or not an integrated surveillance program is justified. Tne j
criteria are:
J l
1.
There must be substantial advantages to be gained.
such as reduced power outages or reduced personnel exposure to radiation, as a
direct result of not requiring surveillance capsules in all reactors in the set.
j 2.
The design and operating features of the reactors in the set must be sufficiently similar to permit accurate comparisons of the predicted amount of radiation damage as a function of total power output.
j 3.
There must be an adequate dosimetry program for each i
reactor.
]
4.
There must be a contingency plan to assure that the
]
surveillance program for each reactor will not be jeopard 12ed j
by operation at reduced power level or by an extended outage of another reactor from which data are expected.
5.
No reduction in the requirements for number of
]
l materials to be irradiated, specimen
- type, or number of L
specimens per reactor is permitted, but the amount of testing l
may be reduced if the initial results agree with predictions.
j l
1 6.
There must be adequate arrangement for data sharing between plants, l
L l
Turkey point Units 3 and 4 began operation with three capsules l
containing weld metal specimens in each of the Turkey Point Units -
1 1 !
l
t I
i
+,
one of capsule T,
one of capssle V,
and one of capsule X.
f Intervenors have already demonstrated that when the first weld metal capsule T
specimens were tested in 1976'in order to revise the pressure / temperature
- limits, the tests showed that the Unit 4 weld metal was found to be the limiting material for controlling the vessel RTNDT because it exhibited a greater sensitivity to neutron radiation embrittlement in that there was about a 30% difference in calculated RTNDT, (Exhibit 6 at 27).
The SWRI report on the testing of capsule 4 also suggested that "because of the potential of reaching a
low Cv shelf energy condition in the Turkey Point Unit 4 weld metal in the ne)t few
- years, it is advisable to obtain another data point in the not to distant future. (Exhibit 7 at 36).
Another report by SWRI dated May 1979 and entitled Reaqtgr Vessel Material Surveillance ProorgO suggested that capsule V be removed from each unit after approximately 7 EFPY of operation and that the data obtained from capsule V be used to revise the P/T limits beyond 10 EFPY. (Exhibit 12).
According to the Licensee's surveillance program in existence at that time capsule V from both Unit 3 and 4 was scheduled to be removed from both units 3 and 4 and be tested on or about 1985.
- However, in February 1985 the Licensee requested and was later granted a
license amendment which allowed them to integrate their surveillance programs for Units 3 and 4 and delayed the test of the Unit 4 capsule V surveillance specimens until 1997. (Exhibit 13). _
i Intervenors contend that the Licensee was improperly and i
i perhaps illegally granted this license amendment by the Staff 1
because the Licensee did not meet the criteria of an Integrated Surveillance Program when the amesidment was granted', and they still do not meet'these requirements.
I First of
- all, the Appendix H criteria states under Section II.C(5) that the testing may be reduced if the initial results agree with the predictions.
The documents presented herein prove that the f
test results for Unit 4 did not agree with the predictions.
This view is supported by Licensee's response to Intervenors' Interrogatory B.4 where they state that the adjusted reference l
temperature for Unit 4
capsule T,
the only tested capsule, was higher than the adjusted reference temperature predicted by Revision 1
Furthermore, the Affidavit of Stephen A.
Collard (September 11, 1989) at 46 states that FPL had informed the NRC on several occassions prior to the Staff issuance of the Safety Evaluation on the amendments of the discrepancy in the test r
f results for the weld capsules from Turkey Point Units 3 and 4.
The Staff objected to answering Intervenors' Interrogatory No.15, which asked them why they allowed FPL to implement the Integrated Surveillance Program when results for Unit 4 capsule T did not agree with predictions, j
- Second, the Appendix H criteria requires that the design and operating features of the reactors in a set must be sufficiently similar to permit accurate comparisons of the predicted amount of L
l l l l
l 1
I h
i radiation damage as a function of total power output. Intervenors contend that at the time they were permitted to implement the i
Integrated Surveillance Program the Licensee and th'e Staff reali:ec i
that implementation of the flux reduction program designed to cut down on the amount of neutron irradiation bombarding the vessel walls, would mean that Turkey Point Units 3 and 4 would be operating with mixed fuel cores that were not identical in nature, and that this practice continues to date.
According to an NRC document dated February 27, 1965, re: "Near Turkey Point Plant Units 3 and 4"
,the flux Term Flux Reduction reduction program was implemented for cycle 8 in Unit 3 and cycle 9 in Unit 4 (Exhibit 14 p.4).
A review of an FPL document entitled Eg_atclat_CJyity_Nevtrqa Measurtment Procram for FPL Turkey Point Unit 3, datea April 1986, states "over the lifetime of a nuclear power plant, changing fuel management schemes can result in significant changes in both magnitude and distribution of neutron flux and hence, neutron fluence throughout the reactor vessel beltline region." (Exhibit 15, pp.1-1 to 1-2).
A review of the reload Safety Evaluation documents for Unit 2, cycle 10, and Unit 4,
cycle 10, demonstrate that the units were operating in cycle 9 with different fuel core mixes.
For example, Unit 3 was operating in cycle 9 with 56 Westinghouse optimized fuel assemblies and 101 Westinghouse 15 X 15 low parasitic (LOPAR) fuel assemblies.
(Exhibit 16).
Unit 4 was operating in cycle 9 with 611 _. _ _
l t
Westinghouse 15 x 15 low parasitic (LOPAR) fuel assemblies. (Exhibit 17).
The differences in fuel were continued in cycle 10 with Unit 3 operating with 112 Westinghouse optimized fuel assemblies and 45 Westinghouse 15 X 15 low parasitic (LOPAR) fuel assemblies, and Unit 4
with 117 Westinghouse 15 X
15 low parasitic (LOPAR) fuel assemblies and 40 Westinghouse 15 x 15 optimirad fuel assembliss.
(Exhibits 18 and 19).
It is interesting to note that on page 4 of the Staff's Safety Evaluation attached to the 1965 amendment granting the Integratec Surveillance Program it states that, "If future core designs are significantly different from those documented by the 1.1censee, the Licensee must explain the effect the changes have on neutron irradiation damage and the surveillance capsule withcrawal schedule." (Exhibit 9, p.4).
It is incomprehensible to Intervenors why the Safety Evaluation, which did not document the discrepencies in the Unit 3 and Unit 4
capsule results also does not document mixed fuel core design changes that existed at the time of issuance of the amendment e
and to Intervenors' best belief will continue to exist until Turkey Point Units 3 and 4 have achieved homogeneous cores some time in the future.
Intervenors
- contend, as does Dr. George Sih in his letter of
'l October 10, 1985, that " loading history" plays a major role in the embrittlement process. (Exhibits 11, p.2).
- Third, Intervenors also contend that the Turkey Point units have had marked differences in capacity factors in some years that could jeopardi:e the integrated surveillance program.
Stepnen Collard testifies in his affidavit at 54 that if one unit has an f
extended outage or period of low power operation the test data from the unit which experienced the extended outage or period of low power operation could correspond to a relatively low fluence and might not be suffeient to confirm the existing fracture toughness of the reactor vessel of the other unit.
(Collard af fidavit at 54 ).
It is interesting to note that the capacity factors for units 3 and 4
in 1984 the year before the amendment was granted were significantly difforent.
In 1984, Unit 4 had a capacity factor of 81.0
% and Unit 3 of 52.6%.
Even more striking is the fact that in
- 1981, Unit 4 operated at a high 78.5% capacity factor, while Unit 3 operated at a mere 16.1%. (Exhibit 20).
These divergent capacity factors continued to exist subsequent to 1995.
According to Licensee's
Response
to Intervenors' Interrogatory B-1, in 1985 Unit 3 had a 75.9% capacity factor and Unit 4
a capacity factor of 29.7%. In 1987, Unit 3 had a capacity factor of 15.3% and Unit 4 of 45.1%.
- Fourth, despite these differences in capacity factors and although Stephen Collard states in his affidavit at 55 that Appendix H
to 10 C.F.R. Part 50 require each integrated surveillance program to have a
contingency plan to ensure that if one unit in an integrated surveillance program has an extenced outage or period of low power operation, surveillance capsule test data will be.
r b
i i
l l
available with fluences comparable to the fluences being accumulateQ f
by the other operating units in the integrated program. Intervenors L
contend that the Licensee does not have an adequate contingency plan to ensure that these differences in capacity factors will not
+
compromise the program.
Intervenors contend that this is so because in response to i
Intervenors' Interrogatory B.3 which asked for icontification of the contingency
- plan, Licensee referred Intervenors to documents supplied to the NRC on February 8, 1985, and March 6, 1985 as part of their amendment request.
A review of these documents suggests that FPL did not then, nor do they now, have a concrete contingency plan to meet the requirements of Appendix H.
For instance in the Safety Evaluation attached to the Licensee's February 1985 letter under Continoency Plan in the Event of Reggced Power OD31ptions or Extended Outace it states: "Both plants have capsules." (Exhibit 21 SE p.2) (Also, see Collard Affidavit at 49).
Additionally, when Intervenors reviewed documents produced by t
Licensee in response to Intervenors' document reQuett, they were advised by counsel for the Licensee, John Butler, that there was no 3
9 written document entitled " Contingency Plan".
Intervenors contend that the Licensee's failure to have a contingency plan to ensure that they are correctly calculating the fluence to the vessel and subsequent reduction in fracture toughness means that they do not now nor have they ever met the requirements of the integrated surveillance program.
~22-
i l
l I
Accitionally, in relation to differences in capacity factors, Intervenors would like to address Licensee's spurious argument that even if a difference in capacity factors or'EFPY were postulated to occur since
- 1985, and even though it would be possible for the remaining capsules in one of the Turkey Point Units to have significantly less fluence that the fluence of the reactor vessel of f
the other unit, such a result would only affect the ability to make f
f predictions or extrapolations beyond 20 EFPY, since the existing surveillance data are sufficient for predictions or calculations up to 20 EFPY. (Collard affidavit at 58).
I i
Intervenors contend that this argument is not correct, since 1 #
the fracture toughness of one unit is being compromised by the other
- unit, which has had a period of low operation, it would be prudent, say in the case of Unit 4 which has not been tested since 1976. tc test capsule V
to assure that the P/T limits are conservative.
Stephen Collard himself states that schedules for removal and testing of surveillance capsules are designed to confirm the existino fracture touch _ngga of the reactor vessel as well as to make predictions. (Collard Affidavit at 53). (Emohaghis SypplieQ).
Additionally, Intervenors would also like to take issue with an argument that has been used for years to allow the Licensee to use an integrated program to predict radiation damage to Unit 4 That argument is the one used by Stephen Collard at 38-43 of his affidavit and by the NRC Staff on page 7
of the NRC Safety Evaluation Quoting Prior
- Rancall, which apparently attempts to
~23-
E I
justify the integrated surveillance program and discount the 1976 l
weld metal test results for Unit 4 by attributing the high test l
result to the alleged difference in flux lot number for the sample in Unit 4.
[
l Intervenors have seen this argument used numerous times as a reason why capsule T for Unit 4 may have tested so much higher than Unit 3.
In fact, this argument was first used by the Licensee in a letter to the NRC dated April 11, 1977, one year after the SWR 1 Unit 4
capsule T test results documented in the first part of this brief demonstrated that the weld metal Unit 4
was already highly embrittled.
In their 1977 letter to the NRC, the Licensee states
Howe ve r,
the weldment samples for Unit 4 surveillance capsule T, t
although containing the same filled wire heat
- number, useo a d1fferent weloing flux lot number. Therefore, the Unit 3 capsulo T sample is more representative of the Unit 4
reactor vessel."
(Exhibit 21).
Intervenors have documented the fact that the Licensee used the "more representative" argument to justify using Unit 3 data for Unit 4
in response to the NRC's 10 C.F.R. 50.54 letter regarding pressurized thermal shock concerns relative to Unit 4 well before the Integrated Surveillance Program was granted. (Exhibit 22).
Yet, in response to Intervenors' Interrogatory nos. 7 and 6 the i
Staff responds that flux lot is oniv of minor imoortantp in determining the sensitivity to irradiation embrittlement. ( Kmp_hgElg l
1.MaplitL.)
l l
l
l If the Staff is correct in their statement, does this mean that the damaging test result for Turkey Point Unit 4
is really representative of the damage to the reactor vessel welds, and if it is representative does this mean that the public health and safety is being jeopardized because pressure / temperature limits have oeen non-conservatively set based on the less restrictive Unit 3 data ?
The inconsistencies on this issue alone are simply to important for l
this Board to ignore. Especially in light of the fact that Unit 4 I
suffered from two serious overpressurization events in 1961 which could have caused undetectable flaws in the vessel making it more prone to brittle fracture when stressec. (Exhibit 23).
For all the above
- reasons, Intervenors contend that the Licensee does not now, nor have they ever niet the requirements of the integrated surveillance program identified in Appendix H of 10 C.F.R. Part 50.
Intervenors further contend that because tne Licensee does not meet the requirements of the program, this Board i
should require them to set the pressure / temperature limits for Unit i
4 based on test results from the most limiting material. The most conservative way to accomplish this would be to require the Licensee 1
l to immediately test capsule V of Unit 4 and use Unit 4 capsule T and 1
or V surveillance specimen data to adjust the reference temperature l
and revise the P/T limits for Unit 4.
l l
An alternative would be for the Licensee to calculate the ART l
l and revise the P/T limits for Unit 4 based on only Unit 4 capsule T data but using Regulatory Guide 1.99, Revision 1.
Intervenors ash this Board to reject the Licensee's argument presented at 74 of the r
Collard affidavit where it states that there would be little difference in the P/T limits for Unit 4 if only Unit 4 data was used.
First of all, this curve was calculated roughly on a desk top computer for the purpose of the settlement discussion held between Intervenors and the Licensee.
Second, neither the calculation nor the software program utilized in the detemination of the calculation have been verified by the NRC Staff or any other independent body, such as Westinghouse.
Third, Collard himself states at 75 that it woulc be inappropriate to calculate P/T limits using only one surveillance data point for Unit 4, because such an approach would be inconsistent with Regulatory Guide 1.99.
Yet, in the prior paragraph, he asks the Board to accept this exact type of hypothetical calculation as a reason for accepting the Licensee's assertion that using Unit 4 plant specif ic data would have little effect on the P/T limits.
Furthermore, Intervenors disagree with Mr. Collard. Intervenors contend that in the event that this Board does not agree that Unit 4 capsule V should be tested, it would be more conservative and proper to use the one Unit 4 data point and the Regulatory Guide 1.99, Revision 1 to calculate the ART and revise the P/T limits instead of Revision 2,
which the Licensee used in their hypothetical calculation.
Intervenors also ask this Board to take note of the fact statec at 47 of the Collard affidavit that Turkey Point is atypical among
~26~
plants with NRC accepted integrated surveillance programs in that most of the plants involved in such programs do not have surveillance capsules in their reactor vessels.
- Thus, one can understand the need for such a program for units that have no test i
- capsules, but it is hard to justify such a program for Turkey Point Unit 4 which has its own test capsules, and whose initial weld tests have indicated there may be a high degree of embrittlement.
Intervenors do not believe the Licensee's argument that the integrated program will save radiation to workers meets the "tnere must be a substantial advantage to be gained" criterion of Appeno1x H.
Especially in light of the fact that if all the capsules in both units are to be withdrawn over the lifetime of the units, there would be no cose savings.
Tne dose would merely be spread out over time.
3.
Letter of Dr.
Georce Sih Concernino Issues Relatino to Intervenor's Cont 3ntioii 2.
In a
letter to Intervenors dated October 18, 1989, Dr. George Sih stated:
"...the unit 3 data are incomplete and not sufficient to predict the P/T limits for unit 4 Additional factors such as strain rate and load-history dependent damage accumulation shoulo be considered; they cannot be discussed on an ad hoc basis without analytical and/or experimental support."
"While the P/T limits depend on the combined effects of I
material properties, operating temperature and neutron irraolation as mentioned on p.7, change in strain rate can significantly affect l
the fracture toughness anc RTu o t. This influence has not been taken into account in cetermining the P/T limits."
"No confidence can be placed in determining P/T limits unless the influence of local strain rates on the fracture toughness of reactor vessel materials is accounted for or shown to be otherwise... Damage accumulation is a
highly nonlinear process.
Predictions based on the linear sum is not always conservative...In
- general, Turkey Point Unit 3 and 4 co c1ffer in their loac history.
The information supplied by the ISP 1s not sufficient to concluce that the unit 3 data could be used to predict the behavior of unit 4"
(Sih Letter, Attachment A).
In his letter to Intervenor, Joette Lorion, Dr. Sih takes issue with a
number of Licensee's assertions. First he takes issue with Licensee's supporting argument for measuring fracture toughness describec on page 7-9 in that he states that fracture toughness is strain rate dependent and cannot be adequately described by the work done in ft-1be. (Sih letter at 1).
- Second, Dr.
Sin states in a footnote on page 2 of his letter that Licensee's statement on page 14 of their motion for Summary Disposition where they state that "the rates or duration of accumulation" are not important in considering the effects of neutron irradiation appears to be in contrast with one of the most important unit nyt for measuring 1rradiation damage of materials.
i (1d at 2).
2
- Third, on the same
- cage, Dr.
Sih states that it is not )
f i
1 l
L l
I L
sufficient to draw conclusions from the differences in neutron l
l fluence based on the total sum because material degradation caused by neutron irradiation being accumulative is a time history and rate dependent process. (1g2 at 2).
Dr.
Sih further states that damage accumulation is a highly nonlinear process and thus predictions based on a linear sum are not always conservative.
As evidence of this Dr. Sih uses the data supplied to Intervenors in response to Interrogatory C as a case in l
point.
Dr.
Sih points out that although the total operating time between Units 3 and 4 1s only 4.8%, the deviations on a yearly base are enormous.
(id2 at p.3).
Dr. Sih plotted these figures on a graph and showed that Unit 3 behaved very differently from Unit 4 in that it possessed a slow down period. (1g2 at Table 2).
- Finally, Dr.
Sih concluced that Turkey Point Units 3 and 4 co differ in their loading history and that the information supplied by the Integrated Surveillance Program is not sufficient to conclude that the Unit 3 data could be used to predict the behavior of Unit
- 4. (142 at 3).
4.
Other issu_gs for consideration by the Board Intervenors would also like to present other 1ssues for the Board's consideration.
The first issue concerns the fact that though both the Licensee and FPL contend that Unit 3 has more Effective Full Power Years I
(EFPY) than Unit 4,
- t is d1fficult to understand how this could be l
so in light of the fact that according to information proviced i r. i
1
Response
to Intervenor's Interrogatory No. B.1 Turkey Point 4 has i
nearly 10,000 more Effective Full Power Hours (EFPH) than Unit I and a higher lifetime capacity factor.
Intervenors contend that the difference in capacity factor knd EFPY is important because accorcing to Stephen Collard at paragraph 51 of his affidavit, a
change in EFPY or capacity factors might affect the total fluence which could affect the fracture toughness of the vessel.
Intervencrs would caution the Board, however, tnat l
even though these differences in capacity factor and EFPY are important, they are only some among the many factors that should be considered in cetermining the damage to the vessel welds. (See letter of Dr. George Sih, Exhibit 11 p.2 and Sih letter, Attachment A).
The second issue concerns the fact that tne Licensee may be uncerest1 mating the calculated fluence for Turkey Point Unit 4.
In a l
SAtt_tY F,XAlyg ion ReQALQinQ Pro.1ected Values of Materia _1 ProDertiti for Fraq1vre ToyShness f_pr Prott.ction AQainst Pres $urized TI'.prm41
} hock
- Events, attached to a letter from the NRC to FPL dated March 11,
- 19E7, indicates that the Crookhaven National Lab (BNL) calculation for Unit 4's fluence had a
12%
discrepency w.th Licensee's calculations as opposed to a 3t discrepency between the I
Licensce's and BNL's calculations of Unit 3's fivence. Thus, the Licensee could be underestimating the fluence for Unit 4 (Exhibit i
1 l
24).
l-l Additionally, Intervenors would like to suggest that if one l
1 l l
g.
f i
,N i
'i considers the difference of fluence between capsule T from Units 2 and 4'and then asso,ciates this difference in fluence porportionally
~
to Unit 3. capsule V,
one can predict the fluence value of Unit 4 capsule V.
Fluence of Unit 4 capsule T = 6.05 X 10te Flupnce of Unit 3 capsule T = 5.68'X 10te Capsule Fluence Difference
- 0.37 X 10 s (3.7 X 1058/100) =.037 X 1018 Fluence of Unit 3 capsule V = 1.229 x 1018
- 0.370 X 1018 CAP.gule Fluence Diffgngnce Fluence of Unit 4 capsule V = 1.599 X 1018 Intervenors' believe that the above predictec fluence of Unit 4 capsule V
would produce an unacceptable P/T curve outside of conservative margins of safety embraced within tne acceptable operating -parameters.for operation of the Turkey Point Unit 4 up to i
20
- EFPY, And well above the 1.26 X
lots n/cmr that has been predictec. fo,- 20 EFPY. (Collard Affidavit at 57).
l-CONCLUSION It 1s evident from the issues raised herein, that Intervenors have established that there are substantial and material issues of fact' concerning Contention 2 and that these important safety issues deserve to be resolved at a public hearing.
Set in
- context, the available facts presented to this Board reinforce Intervenors claims that the NRC Staff and FPL have acted improperly throughout the years to avoid, rather than to confront the crucial problem of reactor pressure vessel embrittlement in --
L t
Turkey Point Unit 4 - a problem that threatens the health anc safety of all;who live in the south Florida area.
For all the above stated reasons, Intervenors request that this
~ Board deny Licensee's motion for Summary Disposition of Intervenors Contention 2
and take immediate steps to investigate Intervencr's claims through a full and formal public hearing.
Intervenors would also ask that this Board revoke the subject license amendments at once because the Licensee does not meet the requirements of the Integrated Surveillance Program, the data frcm which served as-a basis for the pressure / temperature limits
-established by the amendments.
k Dated this 19th day of October 1989 in Miami, Florida.
Respectfully submitted, W L%
Joette Lorion, Director L
Center for Nuclear Responsibility I
7210 Red Road #217 Miami, Florida 33143 L
(305) 661-2165 l
l l
l l
i t
j, UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'89 00i 23 P4 :33 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD g-In the Matter of
)
- O " i.. ?.,
)
Docket Nos. 50-250 OLA FLORIDA POWER & LIGHT CO.
)
50-251 OLA
)
Un s 3 and (Pressure / Temperature Amendments)
CERTIFICATE OF SERVICE I hereby certify that copies of "Intervenors' Statement of Material l
Facts As to Which There Is A Genuine Issuc To Be Heard" and i
" Intervenors' Response to Licensee's Motion for Summary Disposition of Intervenors' Contentions" with attatched letter of Dr. George Sih and Exhibits.have been served-on the Licensing Board by Federal Express and on the parties by deposit in the U.S. Mail, l
Postage: Prepaid on the date shown below:
l l
Dr. Paul Cotter John T. Butler Atomic Safety & Licensing Board Steel, Hector & Davis U.S. Nuclear Regulatory Commission 4000 SE Financial Center Washington, D.C.
20$55 Miami, Florida 33131 Glenn O. Bright Steven P. Frantz i
l Atomic Safety & Licensing Board Newman & Holtzinger P.'.
C U.S. Nuclear Regulatory Commission 1615 L. Street NW Washington, D.C. 20555 Suite 1000 Washington, DC 20036 Atomic Safety & Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Office of Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Janice Moore LdC b Chr%
Office of General Counsel Joette Lorion U.S. Nuclear Regulatory Commission Director, Center for Washington, D.C. 20555 Nuclear Responsibility 7210 Red Road #217 Miami, Florida 33143 Dated: October 19, 1989 (305) 661-2165
y f-
- LEHIGH ONIVERSITY Attcchm3nt A
~
h Einstitute cf Tracture and Solid Mechanlcr p'
Packard Lab. D!ds, elp L
. fiETHi.EHEM PENNSYlNANIA 18015 Telephone No. (215) 738.tl30 or 4133 L
p Tax. No. (21!)1!3 402a
_a
( s,.... _.. _ _.... _
[
- -
- i G. C. Sih i
Directot
' Fax:
(305).667-3361 October.18,1989 e
h L
- Ms. Joette Lorion Center for Nuclear Responsibility 7210 Red Road. Suite 217 n.
L Miami, Florida 33143
' Td:
Tur:<ey' Point Nuclear Power Plant Integrated Surveillance Program (ISP).
p Dccunent A.
Affidavit cf Stechen A. Collard on Contentions 2 and 3 by FPL,.
CccuTen* B.
Lic'ensee's Resconse to Intervenors' Firrt Set of Di_scovery Re-cuests to i.icensee (Au;ust 7. 1989_),.
Cear ?4. Lorion:
Eased-on the package of documents you mailed me on the Turkey Point _tiuclear
.P:.ter Plant Integrated Surveillance Program. I find that the unit 3 data are in-
- mplete and not sufficient to predict the P/T limits for unit 4.
Additional factors such as strain rate and load-history dependent damage. accumulation should te consicered; they cannot be oiscussed on an ao hoc basis without analytical ancie experinental support.
The following comments refer to documents A and B referenced above.
f
.(
(1) Pressure /Temoerature Linit (Document A - Section IB7, 8 and 9 en pp. 7 i
to 9 incTusive).
l While the P/T limits de:end on the combined effects of material prcoerties,
'l c:erating temperatu*e and neutron irradiation as mentioned on p. 7. cnange in j
Thi.
act.;ic ute i can significantly affect the fracture toughness and ARTt1DT.
i"luen:e has not been taker into account in determining the P/T linits.
}
1 The su;:oorting argument for measuring fracture toughness from the Charpy V-l:
nc::n tests is not conclusive because fracture tcughness is strain rate dependen:
are cannet be adequately described by the work done in ft-lb. Work donc per uni-l f
tire or ft-ib /see is the relevant quantity in determining damage thresholds.
f 79is' is illustrated in Table 1 for the HY-80 casting material.
Note that the fcur l
csses censidered are the same in ft-lb but the applied strain rates are differer;.
f falling through a larger distance 8 ft identified as Can Tne smaller weignt 30 lbf
!? giving rise to e higher strain rate.
Comparing with Case I. a smali increase i-in strain rate by a factor of 1.1 can lead to ainost four (4) times reduction l
fracture toughness (dW/dV)c which is related to K by the relation lc L
1 i
j
-;Ns. Joette 1.orton October 18,1989 T
Influence of Strain Rate on Yield Strength and Fracture Toughness Table 1.
Determined from Three-Point Bent Sp(ecimen as Specified by ASTM
/
E-23 for HY-80 Casting Material.
Ref. G. C, Sin and D. Y. Trou,
" Dynamic Fracture Rate of Charpy V-Notch Specimen", Journal of 189-203, Theoretical and Applied Fracture tiechanics, Vol. S. pp.
c 1986).
L
~ Case No.
Strain Rate Yield Strength Critical Energy Densit ys(ksi)
(dW/dV),
si)
(ft-1b) e(see-1) o f
1 (1 x 240) 70.36 7S.2B 24.46 (2.< 120) 74.00 79.15 15.70 1?I (4 x 60) 74.80 80.02 10.08 IV (3 x 30) 77.36 80.90 6.47 (IN)(1-2v)Kjc gg)c a -
gf 2ir}E The where v and E are, respectively, the Fo:sson's ratio and Young's modulus.
inst ligament of natorial tnat triggers f ast fracture is r '
c The local strain rates in the reactor vessel wnere defects prevail can be and cannot be known unless a two-cimensional, if not three-dir * :sfonal, ncn-hi q':
No confidence can be placed linear elastic-plast:c stress analysis is performed.
in cetermining P/T limits unless the influence of local strain rates en the frac-I ture toughness of reactor vessel materials is accounted for or shown to be other-This effect cannot be cisnissed on an ad hoc basis because it affects the wise.
calculations of ART, JRTNDT' ' 0' (2) Neutron Irradiatien (3ccumen: A - Section II!B 51 to 65 inclusive on
~
pp. TNTaf.
Referring to tne data en reutron fluence (n/cm ) in Tacle 5 on c. 43, it is 2
I7 n/cm2 (life l
nct sufficient to draw cenc'.usions from the difference of 3.6 x 10 tiee) anc 2.37 x 10 b n/cm; '1985-30) between unit 3 and 4 based en :ne' total sum.
. Material degradation causec by neutron irradiation being accumulative is a tine-It would be more informative to investigate histery and rate dependent precess.
the rate
- at whicn the neutron fluence is accumulated in time on monthly or at The materials on p.14 of Licensee's Motion for Summary Disposition of Inter-venors' Contentions state that -- "the rates or duration of accumulation-- "
This statement net important in consicering :he effects of neutron irradiatien.
apcesrs to be in contrast with one of the most important unit nyt for measuring irradiation damage of.r.aterial.
Here, n stands for the numce-of neut-ons per L
the velocity in cm/sec and t the time.
Rate effect is reflected by v and 3
cm, y and duration by t,
{c c' Ms. Joette Lorion October 18, 1989 h
h least yearly basis.
This point will be highlighted in relation to EFPH.
Predictions based on l
Damage accumulation is a highly nonlinear process.The data in Table 5 are not supportive the 44ncat'4um is not always conservative.
F of tne integrated surveillance program.
(3) Annual EFPH (Dacument B - Section on Licensee's Response C on p.11)..
A case in point on the influence of rate effect can be illustrated by'the Although the differ-annusi EFPH data on p.11 which is summarized in Table 2.
Anrual EFPH fer 'TurkE oint Unit 3 and 4 from 1936-88_.
D Tabla 2.
r Vene Unit 3 Unit 4 t Deviations 1 985 5,032.S 7.706.S
' 53.1 1986 6,652.9 2,601.8
- 60.9
.1987 1.344.6 3,950.2
+193.8 6.7 1988 5.176.2 a.828.9 d
+ 4.8 IT'0T/.4 18,206.3 ence in the total operating tine between unit 3 and 4 is only '4.8%, the devis-A graphical representation of the data tions en a yearly basis are enormous. Unit 3 behaved very differently from unit in: Table 2 can be found in Figure 1.
4; it p ssessed a slow down period.
The two curves intersected at P between An overestimate would result to 1986 and'1987 aside from the initial crossing.
the left of P and underestimate to the right of P should the data of unit 3 be I
applied to predict that of unit 4.
The net damage would not add and subtract as in arithmetic.
The
!c general, Turkey Point Unit 3 and 4 do differ in their loed history.
informa: ion supplied oy the ISP is not sufficient to conclude that the unit 3 data c:uld be used te ;redict the behavior of unit 4.
Very sincerely yours,
.dy W,
! f d.L' George C. Sth Professor of Mechanics l
GCS:bd
Enclosure:
Figure 1 l
l l
1 L
r.,
I J
i q
1 f
s s
20
~
o Unit 3 e Unit 4 i
- i-e.
Od "e
P r
k u
10 Slow down period for Unit 3 i
5(_
i 1985 1986 1987 1983 i.
Data reproduced fr0m section (c) on page 11 of Licensee's Response to Intervenors' First Set of Discovery Requests to Licensee (Au-Fi9ure 1.
Docket Nos. 50-250 OLA-4 and 50-251 OLA-4.
gust 7, 1989):
j; Eiography i
of Dr. George C. M. Sih Professor of Mechanics and Director of the Institute of Fracture and Solid Mechanics l.
I.
Dr. Sih is currently Professor of Mechanics and Director of the Institute I
I of Fracture and Solid Mechanics at Lehigh University Bethlehem, Pennsylvania, r
He also holds the appointment of Adjunct Professor at The Hahnemann Medical Col-
'lege and Hospital of Philadelphia since 1972.
He received his B.S. at the Uni-
~
versity of Portland, Oregon,1953; his M.S. at New York University,1957; and e
Ph.D. at Lehigh University,1960; all of these degrees in Mechanical Engineering.
Dr. Sih has engaged in research in the interaction of mechanical defomation and heat flow (1960) supported by the Koppers Foundation, in Fracture Mechanics (1960 and 1961) for the Boeing Company Transport Division and (1962 to 1965) for L
the National Science Foundation, and as a member of the Technical Staff, Bell Telephone Laboratory (Summer 1961).
He has been engaged as Principal Investigator in more than fifty projects at Lehigh University sponsored by the Office of Naval i
Research, Naval Research Laboratory, the National Aeronautics and Space Adminis-2 ration, the Air Force, the Amy, etc., all of which are concerned with opti-mi:ing the use of high performance material with design, a discipline that has been frecuently referred to as " Fracture Mechanics".
Much of his work has been concerned with estimating the remaining life of material and structural components damaged by yielding and/or fracture.
He specializes in developing ccmputer soft-ware for predicting the mechanical behavior of structures and the stability of
(
cbjects moving through fluid media.
His more recent activities are concerned with the influence of moisture and temperature in composite materials, laser i
glaz1ng techniques and non-destructive testing methods involving high-voltage electrophotogracny.
,),
L
L FroE 1953'to 1957 Dr. Sih was enployed by Radio Corporation of America as l
.a project and research engineer.
He worked on the research and development of-l l-input and output devices for the first generation "Bizmark" computer system.
I Among'the significant' patents he obtained were:
f 1.
Adjustable optica1 system for line printing.
[
- 2. ' Automatic magnetic disc printing device for the Xerox process.
i k
In 1957 and 1958, Dr. Sih returned to the academic life and served at the City College of New York as Lecturer in Mechanical Engineering.
He came to e
Lehigh University in 1958 as Instructor in Engineering Mechanics and was appointed Assistant Professor after completion of his doctorate.
From 1065 to 1966 Dr. Sih
. held the position of Visiting Professor in Aeronautics at the California Institute-of Technology and participated in an Air Force research project on the dynamics of track propagation and si:e effects in the fracture of plates.
Dr. Sih assumed in 1970 the duties of Regional Editor, International Journal of Fracture Mechanics, and the responsibilities of soliciting and reviewing papers in the field of Fracture Mechanics.
From 1971 to 1975, he served as an Associate i
Editor of the ASME Journal of Applied Mechanics.
He is also on the Editorial Ad-visory Board of the Journal of Engineering Fracture Mechanics.
He is also Editor-
.in-Chief of an International Journal of Theoretical and Applied Fracture Mechanics.
1 i
'0". Sih is a Fellow of tne American Society of Mechanical Engineers and Honorary Fellow of the International Congress of Fracture.
He is also a founding member of the International Cooperative Fracture Institute, an organization established to-pronote the interchange of ideas and information among active researchers in fracture mechanics.
2
?pr. 51h.ts also a member of the f ollowing societies:
f 1.
Society of Sigma Xi 2.
ASTM Committee E-24 on Fracture Testing cf Materials International Society of Engineering Science 1
i f
'4.
American Society of Civil Engineering f
[
5.
American Society of Mechanical Engineering International Society for the Interaction of Mechanics and Mathematics f
6.
i' l
Dr. Sih is the Editor of three book series.
Seven volumes on the Mechanics f
of Fracture series have been or are about to be published:
i I
- Methods' of Analysis and Solutions to Crack Problems,1973 Volume I Volume II
- Three-Dimensional Crack Problems,1974
.\\
Volume III - Plates and Shells with Cracks, 1976 Volume IV - Elastodyn'amic Crack Problems, 1976
. Volume V
- Stress Analysis of Notch. Problems,1976 l
Volume VI
- Cracks in Composite Materials,1980 Volume VII - Experimental Evaluation of Stress Concentration and Intensity Factors, 1980 The two other series are Fatioue and Fracture:
1 Volume I
- Fatigue and Fracture S. Kocanda, 1978
- Fracture Micromechanics of Polymer Materials, V. S. Kukshenko Volume II and V. P. Tamu:h, 1980 and Engineering Acolication of Fracture Mechanics:
Evaluation of Structural Compo.
Volume I
- Fracture Mechanics Methodology:
nents Integrity, edited by G. C. Sih and L. Faria 3
. Volume II Mixe9 Mode traca Extea0ioa by E. E. Gdsutos Volume 111.- Fracture Mechanics of Concrete:
Material Characterization and Testing, edited by A. Carpinteri and A. Ingraf fea 1
. Volume IV
.- Fracture Mechanics of, Concrete:
Numerical Analysis and
~
. Structural Application by G. C. Sih and A. DiTommaso Volume V
- Bonded, Repair of Aircraft Structure by A. A. Baker and R. Jones
. Volume VI
- Crack Growth and Material Damage in Concrete:
Limit Load and Brittle Fracture by A. Carpinteri Dr. Sih hes also served as principal organizer and editor of proceedings of several conferences:
1.
International Conference on " Dynamic Crack Propagation" (1972), Lehigh-University 2.
International Conference on " Prospects of Fracture Mechanics", (1974),
The Netherlands 3.
Conference on " Linear Fracture Mechanics", (1975), Lehigh University 4
International Conference on " Fracture Mechanics and Technology", (1976),
Hong Kong 5.
14th Annual Meeting of the Society of Engineering Science, (1977), Le-high University 6.
First USA-USSR Symoosium on " Fracture of Composite Materials", (1978),
USSR 7.
International Conference on " Fracture Mechanics in Engineering Applica-tions" (1979), India L.
International Conference on " Analytical and Experimental Fracture Me-chanics", (1980), Italy 9.
International Conference on " Defects and Fracture", (1980), Poland 4
- 10.. international Conference on " Mixed Mode Cract Propagation", (190,0),
Greeee
.s.
11.
International Conference on " Absorbed Energy and/or Specific Strain En-ergy Density Criterion", (1980), Hungary 12.
International Conference on " Defects, Fracture and Fatigue", (1982),
l l
Canada 13.
International Conference on " Fracture Mechanics Technology Applied to Material Evaluation and Structure Design", (1982)
Australia i
la.
International Conference on " Application of Fracture Mechanics to Ma-l terials and Structures", (1983), Germany t
-Dr. Sih has approximately two hundred publications principally in the area of solid and fracture mechanics.
He has authored and co-authored a total of three books.
l '.
Handbook of Stress Intensity Factors, 1973 Three Dimensional Crack Problems (with M. K. Kassir),1974 l
2.
3.
Cracks in Composite Materials (with E. P. Chen),1980 Dr. Sih received the 1975 Achievement Award from the Chinese Institute of Engineers in the United States and the 1954 Achievement Award from the Chinese Engineers and Scientists Association of Southern California for his accomplishments in research ant' teaching in fracture and solid mechanics.
I Dr. Sih has also been active in serving as members of nations 1 comittees.
Among them are the National Materials Advisory Board concerning with thr Dyn Response of Materials Subjected to High Strain Rate Loading; Ship Materials Fab rication anc Inspection; anc other com.ittees concerning Nuclear Reactor Compo-nents.
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- By Demetrios L Basdekas N.
gasennetag HhaHhaa'f that someday
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. WASEINGTON - There is a high, f,.
. teen, during a enemingly minor anal.
ftmotion ct any of a denen or mon no.
N elser plants around the United States,
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the steal vennel that houses the redlo.
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'acetve care ig,gunas to crack like a
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piece of glass. me neult will be a core
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moltdown, the most serlaus kind of ac.
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Udent.wtuch will injure many people,
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destroy the plast, and probably do.'
M.7 f.'.\\
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E t-55
. stavy the tucJestindustry withit.
- On the thir11 anniversary of tla a..W\\
F 9 "ll'" tree Mlle Island arcadent, the Gov.
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ernment and ladustry are unable or
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towilling to deal bonesey and ur.
g igently with far reaching nuclear.
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- safery problems.
b Archer eenous accident is very
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likely ha'*** the wrces metal was e __
lemed h the reactor vessels, and with
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each day of operatim, neutrtm radia.
- tion is snaking the metal more brittle,
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and more prone to crack in case of
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==:s euidden taunparature change under Prsesure.
- -One manufacturer of nuclear rene.
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' tors has reported to the NucJear Regu.
4:~.
llatory r'emamiamcm that to three to five spore yeen, the vessels in somg gdants wit' he too brittle to operate worker dropped a small light bulb into not thongts vttal to the safe operstlen unfely..But this estimate is wishful an instrument panel, causing an elec.
et a plant ended up canstag earlaus 4htaking, based on unrealisce as.
- trical abort carmat.The short wreaked
- problems, munptions about plant operstors' ac.
havoc on the plant's control systems Tbs Nuclear Regulatory Commis.
tions and accident sequ==== Some
- a variory o' instruments that run sie is charged with ensuring that ab.
Cants are already too dangerous to. crucial pumps and valves - and the cient plants are operated "with eds.
'sperate without cornettve measures, result was that too much water was quate preaa'*m" of the pubbc health l.The commission could do a great pumped through the rinctor, *Hmg and safety. But burinucratic tout.
deal to prevent web an acadent, and it suddenly. It is very doubtful that draggtng and pr=rwmpadon with pub.
strusch out the Itves of many of these *some of the older plants operating lic raianons and financial problems of brittle vessels. if it ordered ?de 'ype of today would be able to withstand the the ind' are contributing to a corrective steps already taken at name abock. Fortunately, Rancho view that technical some European reactors. But the Seco bad been in operation less than lems can watt or do not esdst.
carm.rmoc, regulaung an indwtry two years; tad it been in ope auce for members of the sta*f acknowi.
that has sadous f.sanc1 : and techns.
- 10. tu pressure vesse! most likely edge the safety problems amamated -
enj problems,instead of takmg trJua.
would have ruptured.
with control rystems, but the agency tives tends te sweep difficult tschnical The kinds of cretrol systems that has yet to demand from utilities oper.
problems under the rug remet.'.c3 to want haywire at Ranche Sece are very atmg nuclear. power placts the technf.
cr.'sen only after they occur.
11hely to fall at crucal times to other ca] data on cratml systems ammana ry
..The comminaion must reshze that nuclear-power plants. Whec a pipe to masses the systems' safety this crtsts is upon us. A ternperature bursts, or a seaJ falls, or a valve
- fully, change severe enough to cracA a brit.
sucks, automatic contro: and safery It may be that we need nuclear D reactor vsesel already has oc.
systems almost instantly take acuan power to matotain our standa rd of liv.
surred. in Californsa. but not at one of to compensate, but they do not ajways ing. But then is a vast difference be.
the older, mon vulnerable planu.The take the right acuan.
tween hartag to accept something, commercal nuclemt indust y's ad.
Contro! rvstems are not reviewed and mahne it acceptable. We can mirable safety recort: - no deaths by the Nucient Regulatory Couunis.
make nuclear power acceptable.
Oused by rud2auon - still is intact, sion. They an not immune to fire or The Nuclear Regulatory Commis.
but this cannot last touch longer, be.. power failun; they often have no man chairman, Nunzto Palindina, has Ouse the reactor vesseis and other backups, so are prone to simple fall.
spokse of cleanmg up our nucJear act.
Stical components are agtag.
ure. They are not even ear *bquake.
As a private crtuen, I hope that we do
- For many yean, it has bec known
- Proof, so, bestanmg with virance at the that veneels are becommg brittje.
The N.R.C. rtaff has taken the posi.
N.R.C. One more accident the size of What makes the problets urgent is tian that if a plant gets into trouble be.
Three M1je Ialand's, and the public's that the metal is agmg tnore rapidly cause of contrul system malfuncuana, rescuan almost certainly will fore.
. ~.. -[Ag.R than Expected, and the circumstances it has safety systems to take can of close the nuclear opuan.
that would cause such an accdent now asty problems. But thLs is not so, as Md seem amore likeJy.
events of the last few years show. At Demetnos L. Basdehcu is a twoctor 31.E
, At the Itancho Seco plant, near Sac.
Rancho Seco, at Three Mile taland, safety engineer with the Nuclear 3 v. _., ; 4 rannesen, Call!.. tn Maren lin8 a and at other p& ansa, control systems Regulatory Commission.
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Could Cooling water cracit like a piece of glass. The result an emergency could cause a meltdown
"' ll.b,e a core meltdown, the most sers. instead of preventing one.The cause:
rupture brittle reactor ous esnd ofoccident, u;hkh ueid injure abrupt changes in reactor pressure
+
.I
'f w_ Bll37 geYe are the facts many people, destroy the plant. and and temperature ~a condition called probably destroy the nuclear industrv pressurited thermal shock-would
. By EDWARD EDELSON with it.*-Demetnos L. Basdekas. crack bnttle vessels, allowing emer.
NWING BY EUGENE THOMPSON The New Yorit Times $ larch 29.
gency water to escape.
1992.
The safety engmeer's " piece.of.
.j Basdekas, a reactor safety engneer glass
- charge quickly focused atten.
,l hefe rs o high increessng likelihood with the Nuclear Regulatory Com-tion on thermal shock:
7
.;.at somec,ay soon. during a seemanc-mission, continued his article to warn e The NRC commisssoners held a
"88Mor malfunction at any of a dosen tnat radiation is making the metal re-public meeting.
actor vessols at some nuclear plants e Rep. Ed $1arkey of Stassachu.
more nuclear plants around the
{nited Sta:es, the steel tessel that arttile. As a result. he wrote water setts caHed. congressional hearirm 5
Conenn ued
' **^es the radioactwe cor* is going :n vea to %c ano coni reactor cores tn A*.a isma. is 1
o Werb. en whit ces suppceed to be speedup of ernbnttlemsnt becsuae of joy walls of reactor vessels act:ss the i desnitive study of th) therm 21 the presence cf copper, n:t th1 results country? Reactor vess51 m:nufactur.
pock issue was acc& rated by the of the stand::rd Charpy testa on es.
ers and utilities b2gan leafing
' RC, posed metal samples. This tech-through old 61es to fmd what informa.
V And the kind of debate that has be.
nique-gradually changing metal tion they had about the copper con.
ome quite familiar in recent years temperatures and messunng how tent of metals in reactors.
w s predictably erupted. Electrical much hammer energy the metal can Records showed that there was adulities. reactor manufacturers, and absorb without breaking-actually some copper in the vessel walls them.
the Nuclear Regulatory Commission testa radiation damage. Radiation selves. "We used a lot of auto stock,"
ny that the pressunted thermal. tends to make all metals bnttle;irra.
,xplain,d Marston. "When you melt shock problem is well in hand and disted metal must be raised to a high.
it, you can't get all the winns out."
4 that the " piece of glass" charge is ab-er temperature before it will become But welds m vessel walls were the I
surd. Cntics say that the nuclear peo. ductile. This shift in the transition real problem. Before the industry re.
f ple are talking through their hats be-temperature from bnttle to ductile is glised what was happening, which
/
cause there simply isn't enough infor-a measure of radiation damage.
- was about 1972, spools of copper coat.
3 mation available to assess the danger Nuclear researchers, aware of met.
ed welding wire were routinely used of pressuri:ed thermal shock, al embrittlement, had earlier exposed for these welds. "The copper was used I've recently talked to experts on samples to intense radiation. But the to prevent rust," noted Stephen H.
'}
both sides of the question. At the mo.
surge of reactor construction begin.
Hanauer, director of safety technolo-
.t ment there are no pat answen. But ning in the 1960s found engineers gy at the NRC. "Someone probably a
- nformation about the hazard of ther. without enough reliable data. To an.
got a $10 prize for the suggestion."
j mal shock is accumulating steadily.
Reactor builders switched to nickel.
Here is what you need to know.
coated electrodes, but they couldn't 9
P essunted thermal shock has been replace the welds in older reactors.
udely publicized only recently. But ll Copper was used to When I visited Marston last winter, 3
- nklings of a problem emerged in the prevent rust. Someone the significance of those welds be.
IMOs.
came clear. On his desk was a alab of Y
At one power plant reacter, a work.
probably got a prt2e metal that looked like a paperweight er peered into a video monitor and for the suggestion 33 gone wild. I thought it was eight I
mampulated a robotic arm down into inches wide. But it was really eight 8
- he radioactive water of a 40. foot.
inches thick-the thickness of a reac-high reactor vessel. He slowly 'ished tor. vessel wall. The weld was a yet.
out a small basket hanging near the swer questions about iong. term radia-lowish stnpe in the steel, tapenng thick metal wall of the reactor. Inside tion etTects on metal, baskets of Char.
from three inches thick on one side to the basket was a jumble of pencil size py samples had been positioned in two inches on the other. Marston told steel bars, each alloyed with various early reactors.
me that it can take three weeks of re.
metals and each bearmg a V shaped The pnncipal cause of embnttle-peated passes with electrodes to com-notch.
ment was known to be neutrons, the plete one of those welds. That type of At a nearby test area, he carefully atomic particles emitted by nuclear weld, engineered to be a powerful anloaded his irradiated catch behind fission in the reactor core, colliding bond between huge steel sections of sr.aelded glass windows. Def. maneu. with metal in the reactor. "It's like reactor vsssels, contained enough cop-vers with another robotic arm posi-billiards," says one expert. "Although per to become a. potential ha:.ard twned each steel bar under a wedge-metal atoms are much heavier than instead.
shaped hammer. Then, as samples neutrons, when a high. energy neu.
Interest in reactor. vessel embnttle-were cooled or heated. he pushed a tron collides with a metal atom, the ment heated up in 1977, Marston re.
l button, and the hammer slarnmed neutron forces the atom from its !at-calls. There was trouble with the nto the notches, ucc-the geometne array of atoras "
sample holders in a reactor built by This routine Charpy test inamed The Charpy tests of the 1960s re-Babcock and Wilcox, one of the major far :ts developen yielded expected tri-vealed that just a little copper in a suppliers, he says. Vibration kept Ju!ts. At inwer temperatures, where steel alloy hastens embnttlement-knocking them loose. All the samples metais become bnttle, samples broke Since that time, though, researchers were taken out, and "it looked worse easily. Higher temperatures-like have been uncertain why the pres.
than we thought." Marston said, mdi.
. hose in your kitchen cven-made the ence of copper hastens radiation cam-cating that emonttlement was pro-l steel more ductile. Heated steel sam-age. Theodore U. Marston, who works tressmg faster than expected in the p!es absorbed more hammer energy on thermal shock at the Electric Pow-test samples.
l before snapping, er Research Institute in Palo Alto Added to this continued confirma.
But something unexpected occurred Calif., says there's now strong evi-tion of embnttled. metal samples and when the worker slammed his test dence that neutron bombardment copper contamination of vessels was hammer onto bars alloyed with tiny makes the copper clump together.
an event the following year that, for imounts of ecpper. The steel-even
" Copper starts out in a solid as some, increased the alarm.
warmed-broke easily. He raised the atoms fairly evenly distributed. Un*
On March 20,1978, a worker at the amperature. Still the bnttle bars der radiation the atoms tend to come Rancho Seco nuclear generating plant snapped. Finally at about 300 degrees together as copper particles," he said.
near Sacramerito. Calif., dropped a r, the bars became ductile instead of New mstrurnents that let researcher' light bulb into an instrument panel, bnttle. The presence of copper seemed see atoms within metals show this The panel shorted out and the plant's to be producing strange results. Soon clumping eiTect. Marston says.
instruments went haywire, flashing workers at other power and research As the first discovenes of brittle ir-fake signals to the control systems.
eaetors discovered the name unex.
raciated steel contammg copper be-o.ancho Seco's emergency cooling sys.
, Sed embnttlement.
came known, anxiety began to spread.
- cm nicked mto operation. Cold water W%t puzzled ever.mne was the How much copper was in the steel al-Commed 1
jm
un.nuai in tne mm *sn-e i:rm.
.,auem w m,
. m.... a...... w.
wded mto th) te:ctor, art pp ng t.w
.. mperature from 562 degrees f to an inclusitn of difTerent matenal In a prob:bilistic risk assessment, t
.,5 in a litt!) rn:re than an h:ur.
ta the met:1, an unevenness in the ycu estim:te the lihelihood of en j
- surface, event that initiates a transient, th:n
[,prenure inside the reactor vesselst dropped from the normal 2.200 -
1.'ltrasound inspection is complicat-estimate the likelihood of the reaction J
uunds per square inch to under 1.600 ed somewhat by the fact that reactor to that event, the reaction to that n-i vessels have a Ninch thick clad-action, and so on down the line, bm. Then, as high pressure water
'i ps were tnggered, the pressure ding-s permanently bonded layer-of Westinghouse, for example, has a
-;ent back over 2.000 psi. With no re.
stainless steel on the inside surface computer analysis that) tarts with 17 b,sbie instrumentation to guide them, that can produce false echo patterns. posrible initiators and runs through yntrol room technicians kept the But that's not an insuperable r.ob.
event trees to more than 8,200 end '
aid water nowing, maintaining the lem. Sero says he's impressed 'sy the pointa. The NRC has done the same combination of unexpectedly low tem, sensitivity of the equipnient.
thing. Its numbers come out more or "We've done about a half dozen full. less in agreement about the risk of ersture and high pressure for sever. vespl inspections," Sero said. "You do. thermal shock, But there are inevit 3
i i'
si hours.
The Rancho Seco" transient " as nu-pick up what we call' indications'-as ble differences of opinion about the
.! ear engineers call it, made it clear many as 20 in some vessels. When you a value of those calculations, which
- /9
[ hat pressurized water reactors were pick up any anomalies at all, you ' show that although there is no clear eusceptible to abrupt changes in tem-must look at your pre service inspee. and present danger, corrective action
-[J perature and pressure. Could any tion to see if they existed before and should be taken at some reactors to O
pressurized reactors already have what size they were, reduce the hazard of thermal shock,
'n,3 mall cracks? And could vessel walls "We've found that the equipment Not everyone agrees with the caleu.
' ? ' contaming such cracks, subjected to can pick up things like layers in the lations. 'The NRC may consult its 6
'h sudden changes of temperature and Ouija board and come up with a num-Trf pressure during an acedent, then ber," said Robert Pollard of the Union of Concerned Scientista, "but the er-87 rupture, drainin.t the coolant water and produemg a catastrophic melt-IIThe NRC may consult ror bands on it are so large that it's I. "7 down of the core?
its Ouija board and That's not exactly so, says Chever.
'5S'"ti*l!Y "I "
l.
The truth is that nobody knows for 9et a number, but the ton of 0ak Ridge. "It's ponsible to esti-certain. Calculations indicate that under pressuri:ed. thermal shock con-error bands are 50 mate what the uncertainty in the dmons, a reactor vessel will fail only large, it'S taSeleSS ll analysis is, and you have to live with that uncertamty," he said But you
,2, if cracks of a certain d.mension are take the conservative end of it and 7..
present on the inside wall. Inspec-4
+i t:ons throughout the industry have work with that "
l l "?
ned ultrasound and other nonde-e! adding," Sero continued. "When A lack of data is more or less con-f tructive testing methods and thus we've gone to the inspection reports, ceded all through the NRC report l
E far have found no such eracks. !ndus-we've found that there are layers in "Perhaps the most signi6eant uncer-
?
try representaPves say they are rea-the cladding at the same depth of the tainty m the treatment.
is that i
l '(
onably con 6 dent that no cracks are indication. Our conclusion is that in there are knownlow frequency poten-there. Critics say the inspection all the inspections we've done, we taal over cooling events much more 2
}
-quipment isn't good enough to detect haven't found any indications that we severe than those that have oc-
~'
1 y the cracks. The NRC says its analyses can't resolve as inclusiens of ditTerent curred." the report says at one point.
- i assume that some cracks exist. no material or layers."
"Because these events have not oc-I matter what inspections show.
Sero says Westinghouse gamed cur ed, they have not been taken into t
Richard Cheverton of the Oak con 6dence in the inspection results account in the frequency distribu.
F Ridge National Laboratory, whose when one test showed a gouge on the tion." In other words. it's tough to pre-
'ecm has performed many of the ther-outside wall of a reactor vessel. "We dict the possibility of something that
.][
mal shock analyses, says assump-were able to get pictures of the resetor has never happened. In another see-e hans about weaknesses in nuclear vessel that were taken before it was tion. the report notes " substantial un-
- 1 power plants had to be made. Take metailed." he said. "We found that it certamties* in some estimates and -
l p
'he enucal issue of cracks in the reae-was a gouge that existed before it calculations that are uncertain by 4
iar4essel walls. "It's difScult to look went to the plant /' A sample of a ves.
"plus or minus at least two orders of 4
ar daws after the reactor is m opera-sel wall contaming a crack is used to magnitude, a broad band of uncer-y non, and it's still a question of how calibrate instniments.
tamty, indeed."
y good a job one can do." Cheverton The NRC recently released a de-What else can we do? the NRC peo-
'd)
- aid. "It s not clear yet whether some tailed study on pressurized thermal pie ask. "It isn't well defmed. but it's y
of the shallow Saws that can get us shock and reactor safety. If you really the best mformation we have," said g
into trouble can be found with accura. want a good nght, ask people about the NRC's Hanauer.
t I:
C,7 so we tend to assume that the the reliability of those safety esti-Your best is none too good, the crit-
,y
~aws will be there "
mates. The method the NRC and the ics say. They point out that the prob-e e But Richard J. Sero, who heads a industry uses is called probabilistic abilistic.nsk assessment technique is 1
fr4 ram on thermal shock for Wes. nsk assessment. It's designed to get the same one used in the famous Ras-d
. inghouse 'a major plant builder' around a rather impressive lack of mussen report of 1974, in which a i )
- 'ntains that there is growing evi concrete evidence. All the calcula-team headed by MIT professor Nor-1 $
- nce to support the behef that the tions about pressurized thermal man Rasmussen calculated the nsks geks aren't there. Eng neers often shock, for example, are based on just of nuclear accidents. Rasmussen came e
dk "Wt working reactor vessels with eight events that have occurred at nu-up with some comfortmgly low.r sk
- Unmund equipment. whose echoes c! ear plants. meluding the Rancho fieures. Just last year, though, the
- analy:ed to detect anything Seco transient and the most famous Contwd
. Wet '943 43 c.
3 e-n amewmem
'~~ '
wiu.a va..u v o u.
.u Determinmg a transition tempera. that the nsk.cssessmsnt technio.u, h*RC looked (var the operating estat have cceumul:ted smce then and ture d
' $: ud:d th;t ths odds cf a nucinrent occurr.ng calculated bg Raa6 ceives, a game."
metal, the. amount of radiatizn it re.
World Senes e.fter the first exhibition j
d 85unen were low by a factor of 30, There's also a lot that the utiliti,.
8 stresses to which it is exposed. The and manufacturers can do to leasen ganauer says that nsk calculaton NRC staff used a formula to predict any possible. d s
e learned a lot from Rasmussen's how assumed pre. existing cracks w"cncerity e: Tort. "He kissed off might extend into the vesses wall. say. One easy step is to reshuffle ty;,
','rthquakes in two pages and floods As a tuult of tests on the rate of fuel elements in the reactor een, b
Pi
' e;wo lines." Hanauer noted. Takmg embntilement at various plants, the putting c,lder fuel elementa, which volume of a shelflong safety as.
NRC predicted when some of them emit fewer neutrons, close to the vea.
sessment of the Indian Point reactor will reach a danger point. All things sel wall."!t's easy and cheap to reduce on aest New York City. Hanauer point. considered, the NRC report reached a neutron fhts by a factor of two." ac.
,S ed out that earthquakes and floods reasonably comforting conclusion, it knowledged Hanauer.
s ure toward the top of the list of risks. !!sted 40 pressurized water reactorsfuel elemetta isn't enough. They want Critics.say that,npositioning the A
- he NRC has learned to include such m which pressurized thermal shock p
y.
$sks in ita risk assessments. Hanauer was an issue. "If no one does any. American utilities to reduce neutron
.N.
But Basdekas dismisses the report thing, we've got one reactor that's in exposure g ys.
'd as "the quantification of wishful big trouble, four others that are a lit. dummy fuel t!e behind it, and four that are in a sel wall.Tht.t's been done at two reac.
thinkmg." And George $ih. director of mild kind of trouble." Hanauer told tors in West Germany and one Rus.
A
.?;
the Institute of Fracture and Solidme. "The rest of them will not reach sian. built nactor in Finland. B
,lechanics at Lehigh University. says utilities are reluctant to take the ra.
A
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48 that the impressive report is built on duction in generating capacity that
-(
a foundation of sand.
dummy fuel elements bring.
"The 6amples they study are five There sie many other steps that
!. E menes long, and the vessels are 500 IIThough the inner r.an be take.i, Marston said. one is the l
2 I
.f mehes long,* Sih said. "The sample is portion is brittle, the marvelously simple measure of heat.
serv thin and the vessel is eight outer portion is tough,.
ing the emerg.ncy cooling water to re.
menes thick. We don't know how to radiat, ion damage in the duce thermal shock. Keeping the
. 'J transfer small. sample data to the de.
Wall is attenuated ))
emergency water supply at 120 de.
wn of larEe. scale structural compo.
grees F rather than room tempera.
r.cnts. The scaling e:Tect m size and ture is cheap and effective. Marston aho the scaling eiTect in time are says. Thermal shock can also be re.
amen; the most difficult questions we the screening enterion ;the transition duced by adding controls to throttle a
' avc "
temperaturel.:!unng the anticipated back the automatic.feedwater system,
. if cnties think the NRC has been he notes.
i.
'w peculative. industry believes the life of the plint."
The '* big. trouble' generating plant Improved traimng for reactor oper.
tmart is too conservative. You can a'.
is the H. B. Robmson 2 reactor of Car. ators is another industry option. The l
.c at just about any cene!usion you olina Power and Light. Hanauer cal. idea is to get them ready for all the l
4 l
m.mtwrs. Marston says. "By changin g culated that if noth ng were done. it problems that could lead to a s:gmfi.
- .mt by putting in the appropnate e
cant transient. then avoid the
,e.
assumptions." he explained. "I would reach the transition. tempera.
l an,how that one of these things has eure entenon m September of 1967. quences that end i
- ne
." useful life at all or a lifetime of 30 Turkey Point 3 and 4 m Flonda get The last resort is annealing. The re.
+
'o 40 years." The NRC consistently thee in 19SB: Cab ert Clitis 1 m actor would be shut l
j wo the most conservative numbers Marv!and gets there in 1939: and Fort elements would be removed. :nd the O
Calhoun in Nebraska amves in 1990. vessel would be he ited to SSO degrees Cne of the key factors that the Rancho Seco. Mame Yankee. Oconee F for a week. A study done by Wes.
.t.ts estimates, he says.
n 2 m South Carolina. and Three Mile tinghouse for the Electric Power Re.
t NRC's expens looked at was the tran.
Island I arnve m the 1990s. Every, search Institute concluded that an.
c
" tan, temperature at wnich a piece of thmg else is 2:st century. Hanauer nealing would make the vessel i
=
young agam. The process tsn't cheap.
't.a stops bemg ductile and becomes Reactor rnanufacturers accepted One repon citec costs of 360 rt..llion
~".t e enough to break easily A cru-nys.
those numbers without too much ar.
or more for a single reactor. includmg t
14 pan of the NRC report was to set the pnce of the electnnty that the Pomt at which this transition tem.
d i
gument. "Their conclusions are more or less m line with ours." said Sero of plant did not generate dunng the i
arature in a given reactor would be ye :or concern. The report sets the "mier point at 300 degrees F for ver.
Westinghouse. Sero says that Wes. treatment.
e No one is tSnkmg about annealing l.
fcat welds,270 degrees far horizon.
tinghouse thinks the NRC could set ngnt now. Instead, utilities and man.
3(0"s.
its transition. temperature numbers ufacturrrs are making detailed stud.
about 30 degrees lower but b isn't "igner transition temperatures are fthe ies of all the factors afTecting the ther-
,f5e. Since the reactor vessel must argumg with the basic premiser mal. shock issue for individual plants.
a 1 *namtained at these temperatures report.
The NRC report has asked for such a I
10 dects of bnttle metal are to be Nuclear entics are. They center
=
4 4wided. The origmal standard for nu.
their fire on the vast number of as-plant specific report at least three V
vears before a reactor reaches its
" ear reactors was no more than 200 sumptions that had to be made in the ' screening entenon for danger.
."WS E The temperature :s higher report because information about theFor the Robm>on 2 reactor. the re-2
- d P ical welds becaus pressure probability of difTerent events occur. port would be due in 1964. Carchna nng. nd dout the reiiability of safety mg the possibil
- ty that a erack systems imply isn't available. Rep.
P;wer ano Light ts hard at work, s,iys
~~ 04 'o force the welds nut, mereas.
l
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the radiation damage is attenuated of unknown dangers that lay before D'*$,s' El leman, who is in charge, safety. The vessel wall has through the wall," Cheverton
- them, crack might be dnven through the in.
"The Atomic Energy Commission of nuc leen inspe:ted, and no cracks were ggu training for reactor per, ner part, but it tends to arrest at the went forward with all this undpe opti.
i'8" 't is under way. The egnpany is outer part."
mism," complained Po!!ard, who re.
8'"
But that assessment could easily be signed from his job as a regulator
- g"'.ing a proposal to heat $he emer. wrong, says Pollard of the Union of years ago in disgust. "Now we're in water supply, s'$. utron exposure has been reduced Concerned Scientists. "There's no dis. position where nothing can be done t pute that current emergency systems cor'ect the mistakes without causing W bugung the older fuel elements-the reactor wall. How much would not be able to cope with a frac. someone undue harm. I expected "st ture of the reactor vessel," he said. them to do the job back in the 1960s.
- stra um, will the program buy? "It's mature to speculate about that,"
' Tor other problems, you can make a Now everyone but the nucleastindus-
[,lleman said.
reasonable argument that you have try has to su.Ter."
There's no panic at the NRC, the sorne defense in depth. The defense.
" Sty perception is that the problem manufacturers, or the utilities. The in depth philosophy disappears when is well in hand," said Westinghouse's roblem is well unden,tood, Chever. you talk about pressunted thermal Sero. "We have significant tesearch shock."
programs under way, we are ;autting lon says, and the Oak Ridge analysis ndicated that even if worse came to The real problem. Pollard says, is significant money and engineenng er.
,orst, a reactor vesrel would not that the nation's nuclear regulators forta into it, and we have a firm un.
break wide open. "Even though the and the manufacturers allowed a ma.
derstanding that is going to improve.
.nner portion is bnttle, the outer por- ;c t construction program to roar which will show that our predictions non still is relatively tough because anead without considenng the range were very conservative."
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REFERENCE 5
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April 10,1981 V'.% ; f.,_, ;
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& t.e The Hencrable Morris K. Udall W
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- l. - Chaiman, Subco.mittee on Energy
.N%, s
- 'and the E.1yironment
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/ W+ Comittee on Inter'ior and Insular Affairs 0
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' United States House of Representatives Washington, 9. C. 20515,
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Daar Mr3 Chaiman:
lj
[ Ch May 25,' lh80 iState to you c:ncerning the safety implications of control'
)/f ' systems and. dynamic. characteristics of nuclear power plants. My c:mments then l
were intended to dispute the official NRC position that " safety systems will r.itigate c:ntrol syst.se; failures at any power".
~
j L
One of the specific points I raised then, by way of an example of what Failure t
M:de and Effects Analyses (FMEA's) of control systems' can and should uncover, was the likelihood cf overcooling transients, generated by control system l.
malfunctions in thE sec:ndary side of a pressuri:ed Water Reactor', as described 4
in Reference 7 of _that letter. Such transients can cause the reactor versel to cocl-d:Nn to about 150 *F'in about 15 minutes, while the ECCH repres sur-res it to a out zw psi. Inis como Une transient. k.no.<n as pressur' zed thema9 sheeL 1s : ;a:1e of catastrophically fracturing a ieector vessel that has been nmed I
~
to a neutron fluence c:rresponding to only a few Full Power Years Ecuivalent (rr.rp of operation, e.nc has a high copper content of about 0.4% in itt. walla or weles..,
A reactor vessel fracture is one of the mest serious accidents a reacter r.ay exterience..Cepending on its location and rede, it is almest certain that it' L
will cause a core a:eltdown with all its public health and safety ramifications, L
on.which, ! e:r. sure. I need not elaborate for you.
Considering the high consequences of such an ac:ident, then.one should ask what are the chances H
of.it taking. place. Unfortunately, such an accident is very likely and increa-l
. sin;1y so. It is very likely because it ms,y be caused by one or nere failures in the non-safety c:ntrol systems in the secondary side, and this is substan-tielly supp(rted by operational experience. It is increasingly so because as
'g time ;oes on.the neutron fluence to which the vessels of all reacton are exp sed. is : increasing, and for several of them. I believe that a de.n=er:q1
/~,..
level hn atiendy been reached. I believe that this level is probab1v as_M_
. as 4 rFTE of. operation for vessels with hich cooper 41 1cy walls er waids.
ints is supper:ac. y analyses performed for the NRC, indicating that the over '
E.
cooling transient that took pJace at Ranche Seco on March 20, 1978 would have caused such,'a vessel to rupture, had it been in epcrat' ion for about 10 FPYE.
However, that transient wasinet as severe as we can expect on a reasonable worst case basis. Further:cre, a recent discovery of a discrepancy existing L
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The h norable foi r5 K. U,dall
, April 10, 1981 a
between the estime*id vs'.' the measured values of neutron fluence Faine Yankee reactTr vessel indicates a generic problem that ma.kes things1 werse. The results. of dostmetry measurements indicate th final Safety Analyds Report. Moreover, as you may recall, one of.the measures ordered bf:HRC after the 1141-2 ac:ident was to have all rea l
operators not tur:r off the ECCS once it had been initiated. This might be desirable in some cases of accidents, but not necessarily in of ECCS c:= pounds T$e accident by contributing to the cool-down process, and, most important1y, by repressurt:ing the primary system.
F.
The pressuriied thEhr.a1 shock phenomena have net _ been the subject of expe 2
timental werk by the NRC ner the industry. Hsr nave the control systems.
and their implicat.fons to safety been revimed and analy:ed. These crucial-short::mings pose 's6me questions on the effectiveness of the regulatory process, which you r.ay as easily as I pender, but the imed we are faced todayT'and taking the approach that if we err, we should err in the dire: tion cf* safety, it is apparent to me that' those FWR's with high copper alloy vessels or welds, that have operated for 4 FpYE must be i
shutd:wn until this3 matter is resolved in the technical arena. It is con l.,.
ceivable'that' aftei-addittenal and plant specific studies additional mee-sures ety be"requidd..
i Even thcugh the Comission and the ACRS would probably re.spond to your letter of December:4,1980 on the safety implications of co l
I believe that the Comission, with C ngres-that requires a de:ision new. assistance"and appreciation of the issues involved, sional constructively.
n.7
[
If. I can be of further assis tance, please let me know.
Res p.e:tful ly, r-
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A Demetries L. Basdekas T-Reacter Safety Engineer l
Congressr.an L.uhan s
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Congressman Markey 2
h Chairman Hendrie N
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h NUCLEAR REGULATORY COMMISSION l
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WAsHINo ton, o. C. 20555 i
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August 21, 1981
. w I
.1 Docke t No. ' 50'-251 -
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3 Or. Robert E. Uhr' ig, Vice President Advanced Systems. and Technology Florida Power and Light Company g
P. O. Box 529100 I
's.
Hiami, Florida 33152.
J
Dear Mr. Uhrig:
SUBJECT:
PRESSURIZED THERMAL SHOCX TO REACTOR PRESSURE YES E
We have reviewed the.PWR Owners' Groups responses of May licensees' responses of May 22, 1981 to our letter dated April 20, 1981 15,.1981 and the concerning the subject issue.
The EPRI w t
was included in the licensees' responses.ork which bears on the issue review, ~ of the plants where neutron irradiation has'significantly reducedO 4
the fracture toughness of the reactor pressure vessels (RPVs), all plants could survive a severe overcooling event for at least another year of full power. operation.
taken now to resolve the long-term problems.However, we believe that ad This belief. is based upon our analyses which' indicate that reductions in fractur.e toyqhnees for some RPVs are approachino levels of concern.
- It is also' based in part on the fact that any' proposed corrective action must allow adequate lead time for planning, review, approval, procurement and installation.
These conclusions were recently discussed with the PWR Owners Groups on July 28-30, 1981.. At those meetings, the Owners Groups F
-reviewed the programs underway at the three PWR vendors which are designed to scope the magnitude and appitcability of the generic problem and to be completed by late 1981.
elements for resolution of the problem on a generic basis and th to make full use of the reports due by the end of the year.
While the vendors and Owners Groups are to be commended and encouraged in addressing L
the generic issue, there is also a need for plant-specj fic information for your plant.
Based on current vessel reference temperature and/on system characteristics, we have identified Ft. Calhoun, Robinson 2, San Onofre 1, Maine Yankee,.
Oconee 1 Turkey Point 4, Calvert Cliffs 1 and Three Mile Island 1 as plants from which we require additional information at this time.
The staff has used the time-dependent pressure and temperature data from the March 20, 1978 Rancho Seco transient as a starting point for our evaluation of this issue because:
(1) event experienced to date in an operating plant;it is the most severe overcooling (2) it is a real, as l,
o i
a a
j.
l, Dr. Robert E. Uhrig L L
i- 0 opposed to a postulated, event; and (3) itag enough that it could challenge the PfV when enmhingd ut *b nhysjCally reasonabic values of IP-radiated fracture touchness and h4*b1' eracTstze.
in ruture reviews the 1
staff plaiis to use the steam line break accident or other appropriate transient / accident in order to estimate minimum operational times available before plant modifications are required.
- Using calculated RPY steel mechanical properties, credible initial flaw sizes, reasonable thermal-hydraulic parameters, and a simplified pressure-7 temperature transient similar to that observed during the Rancho Seco i
event, the staff has concluded that all operating plants could safely' survive such an event at the present time and for at least an additional year of full power operation.
However, because of the_recuired lead _ times for future actions, the margins in time for lona term ooeration are not large, ahTthere Is constdcrabie uncertainty in tife probability triat similar
~
or more severe transtents may Uccur.
It is clear that positive action must be i ni tia teLscon..for..those ol ants wi th sitini ficann v nicn trans1 tter-l tempe r_a turg s.
As indicated above, several such plan'ts have beenTelected by the staff, based on estimates of the current reference temperature for the nil ductility transition (RT
) of the RPYs.
NDT nitiate further action at this time is emphasized by the The need t(that igTEmentation of7n)rytposed fixes or remedial actions recognition
\\
must allow foF adecuate lead time.
Because long-tem 7crluthns may require a year or mne, you should explore short-term approaches as well.
Although clear, concise instructions should be provided to operators to reduce the likelihood of repressurization during overcooling transients, the NRC staff believes that reliance on operator actions to prevent repressurization during an overcooling transient will be ver/ difficult to justify as an acceptable long-tem solution to the problem.
In accordance with 10 CFR 50.54(f) of the Commission's regulations, you are requested to submit written statements, signed under oath or affimation, to enable the Commission to detemine whether or not your license should be modi-fled, Suspended or revoked.
Soecifically, you are requested to submit the L
following infomation to the NRC within 60 days from the date of this letter:
l l:
(1) Provide the RT values of the critical welds and plates (or for-l NDT gings) in your vessel for:
l (a) initial (as-built) conditions and location (e.g.,1/4 T) and (b) current conditions (include fluence level) at the RPV inside carbon steel surface.
l l
L 1
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H Dr. Robert E. Uhrig 3
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1 I
(2) At what rate is RT increasing for these welds and plate material?
HOT (3) What value of RT for the critical welds and plate material do NOT you consider appropriate as a limit for continued operation?
(4) What is the basis fer your proposed limit?
(5)
Provide a listing of operator actions which are required for your plant to prevent pressurized themal shock and to ensure vessel i ntegri ty.
Include a description of the circumstances in which these operator actions are required to be taken.
Included in this summary should be the specific pressure, temperature Mid level values for:
a) high pressure injection (HPI) temination criteria presently used at your facility, b) HPI throttling criteria and instruction presently used at your facility and c) criteria for throttling feedwater presently used at your. facility.
For each required operator action give the information available to the operator and the time available for his decision and the required action.
State how each required operator action is incorporated in plant operating procedures and in training and requalification training programs.
A You are also._te_ quested to submit a clan for_ Torby DMa*
Unit Ho, 4 20 the NRC within 150 days of the date of this letter that will dafine actions and s'cTedles for resolution of this issue and analyses supporting.
~
continued cperation.
We request that you include consideration and evalua-i tion 6f7he following possible actions:
/
-(l) reduction of further neutron radiation damage at the beltline by replacement of outer fuel assemblies with dummy assemblies /
'i or other fuel management changes; l
(2) reduction of the themal shock severity by increasing the ECC water temperature; (3) recovery of RPV toughness by in-placa annealing (include the basis for demonstrating that your plant meets the requirements in 10 CFR 50 n
Appendix G IV C);
\\-
(4) design of a control system to mitigate the initial thermal shock and control repressurization.
t
.For these, as well as for any other alternative approaches, provide implementa tion schedules tha t wquid_.assur.e.J;_ontinuance of adequ.a te I
sa fety margins.
~ ~ ~ ~ ~
In the Interest of efficient evaluation of your submittal, we request that you include wi th the above plan, a response to the enclosed reouest for addi tional information.
j l
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Dr. Robert E. Uhrig 4
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i Due to.the nature of this review, and the past review effort that has been expended, we' consider the 'above schedules to be reasnnable; however, infom
.us within 30 days if you anticipate conflicts with previous commitments with either submittal, and a basis for any delay.
We also expect participation
' by the appropri,1te PWR Owners Group and NSSS vendors in developing solutions
)
to the problem.'
Sincerely.
s A4 Oarrell G. Eisenhut, Of rector Division of Licensing Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Infomation cc w/ enclosure:
See next page P
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October 23, 1981 l
L-81-465 l
Office / Muclear Reactor Regulation Attent y.: Mr. Darrell G. Eisenhut. Director
'83 \\ 'S ! g.
Division of Licensing i
U. S.,f ear Regulatdry Comission
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i Octokr 21. 1581 '
gpg49tENTCORRECTION
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gy. our let ter referenced above, we responded to. questions g7 jptter dated August 21. 1981 ift Unip 3 & 4 relating to pressurized themal shock regarding Turkey
,,,,,c., pressure vessels.
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g,g.f.d corrected page replaces Attachment Page 1 L
of our let er referenced above.
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1 Very truly yours,
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Rooert'~. lFIS L
Vi ce P res * **'t Advanced 53'tecs & Technology i
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W. Ma P. 0':.eilly. Region II 2
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Pressurized Thermal Shock to teactor Pressure Vessels t
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QuestionIll:
Values of the critical melds and plates (or forgings) in Provide the RTNDT your vessel tor:
' Initial *(as built) conditions ana loc.ation (e.g. '1/4T) and E'I'(.;A.
currgRt conditions (include fluence level) at the RPV inside carcoa 6ttel a
W b
. ' surface.
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.'.'i';.. N. Res ponse (1):
Initial Current e.
Mat eri al h
RTNOT*
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Intervsdiate Inner 123P481VA1
~~ % 50 F
+ 35 F
+ B5F.
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Forging *
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r Ci rooferential 7.!cs.
- ti (Giru) Weld **
SA 1101
+3F
+190 F
+193 F
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1.ower Forging
- 1225180VA1 40 F
+ 35 F ; + 7 5F 1/4 T The current RTggr 11/4 T) = +168 F.
Value is baseg on
- nner wall.
., nit 3 'Jata which has been shama to be more representative of '.et t 4
- nan surveillance capsule renc=ec f ran Unit 4 '(L-77-113, datec J4ti) '
11,1977 and L.77-326, dated 2.oser 21,1977).
!ased on the slope o'f predict. inn curves presented in proposec A!'M tandarcs "Precicting Neutron aciation Damage To Reactor Vessei
+
t mat e ri al."
There 'tave been 5.61 Ef fective Full Power Years (EFPY) of operation ay of (b)
Sepenser 30, 1981 at Turkey Point mit 4.
The :::al fluence on the inner wall is 1.1 x 1019 n/cm2 and 6.6 K I J n/:a' at 1/a T.
Question '2):
increasing for t:1ese welds and plate material?
At wna: rate is RTND7 Rescoase (2):
- ne rate of 7'F/~7PY f or tne nex: 10 yea rs ; f or ::
Rigg-is tncreasing a:
Tr.e rate :.f :.nange f:r ne f or;ings i s 35 4, reca c:e* of li f e. P~/EFDT.
e cerr.e t
- ces ;n ' it ;f
- ne vessel.
Tr.ese aae oese: on :ne s '. ::>e f
- re:* :; :r :v< es : e s e
- e-: : ao:0:ec U3 Stanca ::
v o :: i n g 'it
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ANALYSIS OF CAPSULE T FROM THE FLORIDA POWER AND LIGHT COMPANY 7
h TURKEY POINT UNIT ~NO. 3 REACTOR YESSEL RADIATION j
SURVEILLANCE PROGRAM l 4..
T S. E. Yendko l<
J. H. Phillips I
S. L. Anderwn j
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December 1975 j
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Wort performed under Shoo Order No. MIP 23572 s
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APPROV E D:
y N.
I Structural Mateculs Engineering WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems F O. Box 355 Pittsburgh Pennsylvana 15230 j,
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- rgit.on is 6 $5 a 101I 2 (E - 1 Val This f!wence is accreter a'ed, the tar.e a em i
- 5 tre 6 3C ICII n,=m fiwence u6cw!atec for 40 vur oce'ation.
l o
- 7...: ::r e,ec t ec eif t in A TND7 of tne weid metal af te 'Oca:e**ssr, car :cretion es l
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f weiy
~
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p O
Tag.r *b:.alrc pro:erties Cf !Orying IUP461V A I and the MC me'.4! are >cecuate 13 I
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SOUTHWEST RESEARCH IN S TITU T E Post Office Dro-er 28510, 8500 Culebro teod l
$en Antonio, f o s os 7 8 28 4 ACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FOR
~
TURKEY POINT UNIT NO.4 ANALYSIS OF CAPSULE T IT E s. Herris l
l i
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FINA L REPORT I
8= R1 Project No. 02 4221 M!t i.
j i Florida Power & Light (:ompan,$
P. O. Bo x 3100 M
Miami, Florida 33101 ij l
,j lj Ju ne 14,19 74 E!
i lS Approwed:
0 I
U. 5. Lindholm, Direct o r i
Depa rtment of Motorials Science 6 I
3 Loc.i t ion len d Shift in RTNDT (dee F1 in Wall Fac.or 3 FFPY
$MTPY 10 F.F P Y 32 ETPY 1/47 4.17 242 24) 342 467 3 /4 T 17.4 162 1:44 230 312 T he s e va lue s,we r e u s e d a s t he ba s e s f o r c omputin g he a tu p a n d c ooldewn i
I limit curves for Turkey Point Unit No. 4.
(Three ETPY wt!! not be r e a c he d un til ne a r the e nd of C o r e C y c le IV a s e s timate d f r om b oth compute r predic tions and pa st ope rating expe rience. )
(i)
As suming that the pe rcent change in Charpy Y. notch upper shelf energy is proportional to the square root of the neutron fluence, the we d metal upper shelf energy at the 1/47 positten is predteted to reach the 50 ft.lb level at approximate!T 2. 7 ETPY of ope ration.
t (101 Althou gh the s u rveillanc e p rogr am is in c omplianc e with Ap.
l pe ndix H of 10CTR50, it is recommended that a replacement capsule with additional weld metal s pecimens be placed in the Capsule T alot if archiva!
4 mate rial is available. An a lte r na tive is t o m o ve Ca pe u te V iate the Cap-s ule T slot at the end of Core Crcle 112 (April 1977) and remove it for tes t.
[
't ls in g a t th e e n d d C o r e C rc le lY I A P r il 19 78 ). a t w hic h time the e s tima te d
,k fla enc e on Ca ps ule Y w ould oc 8. 2 5 z 10 neutrons per em2 (E > ! MeV).
38 11
(!!)
Ce the ba s is of NR C r e c omme nda tion s, e.e W OL f r ac tu r e i
I me c ha n ic s s pe c ime n s ha ve be e n s t o re d u nte s te d pe n din g de ve lo pme nt of r e c omm enda tlas s c onc e rr.in g te s t p r oc e du r e s.
s
,4 l
w d
.l
p-38 1
1 i
have been The refore, the projections of shif t in RTNDT 19
{
g ne a rly I 'x 10 The result obtained from Capsule T has l
y,,4 on the We stinghouse curve s.
7 d response curve has been drawn
..,,n added to Tigure 9, and a normalize The predicted i
h curves.
,3, e.g n the da ta point p a r allel t o the W e s t n g oa s e i
for the Turkey Point Unit No. 4 reactor pressure vessel
.3dts n ETSM The value s predicted j
it em Tigun 9 are summarized in Table XJ.
r
,m,x :
ldown limit curves l
f
,t :nc 1 '4 7 and 3/4 7 a re used to develop heet9p and coo ASM T: Code.
-.,ee the requireme nt. of Appendia e, to S.'etmn !!! nf the f
l-3 s hif ts s helf ene rgy reduction s and RT g T e se proja ctions for Cy i
Capsule T, and trend curven for like materials.
/
a r e sa s e d on one data potat.
ll be improved as
- t is anticipated that the reliability of the trend curves wi i
ding of the more surveillance data become s available and a better unde rsta As an esample
{
factor a affecting radiation embrittlement has been achieved.
of the latte r, Mr. E. C. Biemiller of Combustion Ensinee ring, la a pape r I
l f
l
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given at the ASTM Elo Sympoelum en Effects of Radiation on Strectura 4
Louis, May 4 6, indic ated tha t a ps rame te r of (', Ni + *. SI)
M ate rials in 54.
i l
bi l t
j
- ('. W o + ". C r + ". W n ) ma y e x pla in the va r ia tion in r a dia tion e m r tt eme n
- Also,
- sse rve d in fe rritic materials of nominally the same coppe r content.
l that
}
the Metal Prope rtie s Council is developing new radiation damate curve s r
j l
- ttl be ba s e d on m o re da ta than thos e c u r r e n tly in u s e,
t shelf energy condicion in Secause of the potihtTa! of reaching-a-low-Cr ble
,he Turkey Point Unit No.,4 we!d metal in the next few yearodit:Ic-advis
=
j>
10 ootain another. data point in the not too dist. ant future.
p
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""*"~9"vagm._n%.,p
5OUTHWCST RESEAtCH IN5ft?u?E Post Clfite Drewer 28510.5$00 Celebee Road 5en Antonio, Te se s 7 828 4 L
PRESSURE-TEMPER ATURE LIMITATIONS FOR THE l
TURKEY POINT UNIT NOS. 3 AND 4 l
NUCLEAR POWER PLANTS-
\\
I by E. 8. Not tis J. F. U nt wh
.Nw RI Prnject het. (I2. I.*I8.3.(13 9 to l
Fla rid a Po = c r a nil I.is hi (*.a m pa n,,
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Mia mi. Fleerida June 301976 Approved:
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U. 5. Lindholm, Direc to r Department of Moferials Sciences
26 g
s The first surveillance capwle was removed from Unit No. 4 during the 1975 refuelling outage. This capsule (also identified as "T") was eval.
usted by SwM, and the results have also been reported.* The Unit No. 4 weld metal was also found to be the limiting material for controlling the ves sel RTNDT, and it exhibited an even greater sensitivity to neutr'on re-3 i
diation embrittleme nt, i
As a pa rt of their analys is of Capsule T We stin ghous e' computed g
Scatup and cocidown limit curves for. Unit No. 3'for 3 and 10 effective full
)
powe r yea r s.
In delt analrels, they employed additional conservattam above ASME 5ection 112 Code requirements by appl %ag a 1.25 safety factor to the s tre ss inte nsity factor cau sed by thermal g radients. Florida Power l
4 !Jaht Company asked SwM to recompute the Unit No. 3 heatup and cool.
l f
down limit curve s and comput.2 heatup and cooldown limit curves for Unit
- t 4
No. 4 using the s afety fac ter s called out im Append!.x C of Sec tion III of the i
1 A.5M E Code.
r 1
i B.
In eut Info rma tion i
k j
T r a c tu r e Toug hne s s P r o me r tie s The value s of RTNDT fo r t.se be ltline r e g ions o( Tu r k e y Poirm Unit Nos. 3 a nd 4 w e r e de r ive d f rom ( ! ) the s u rv e illanc e p r o g r a m te s t r e -
i
{
sulta, (2 ) c omputed ratioe of ta s t fluz at the ca psule loc ation to the mAzi-
{
i
.6 mum fast flux at t.he 1/4T and 3/4T locations in the ve s sel walk and l a' i
y
- No r r i s, E.
B.,
' Reactor Ye s s el Ma te ria l Su rv eillanc e P rog ra m for
- ')
Tu rk e y Point Unit No. 4 - Analys is of Ca ps ule T. " Final Rs po r t. SwRJ s
. P r ojec t 02 4221. Jane 14,1976.
lt
17 (3) trend curves of increase in RTgo7 as a function of neutron fluence I
i (g > 1 MeV). A summary of these values is as fo!!ows:
j-I Unit Ope rating RTNDT RT NDT No.
Pe rlod*
at !/4T at 3/4T i
(
3 5 ETPY 194'T 131*T 3
40 ETPY 236'T 159'T 4
5 ETPY 251*T 188*T 4
to ETPY
}42*T 00*T ETFY
- Effectbe T'ull Pe-er Yea r s.
l Ve s s e! Cois tants
{
v, The followin3 nput data were emplayed in this amalysies i
~ if f
Irne r Radius', r i
- 77. 7 5 in.
=
8 5. 7 8 in.
Qa t e r Ra diu s, r.
i 3235 peig
]
Ope rating P r e s su r e.
P, a
?l 70*T Initial Tempe rature, T, s
550'T rinal Tempe rature. Tt 6 lbm / h'
=
97a10 Effective Coolant Flow Rate. Q 9
l Effective Flow Area. A 19.15 ft-
- 11. 9 in.
Effectin Hydraulic Diamete r. D 3
C.
Heatus and Cooth Limit Cu rve s Slace He a tap cu rve s we r er c om pe te d fo r a he a tu p ra te of I Oo
- r / hr.
low e r r a te s te nd to r ai s e the c u rve in the c e nt r al r e g ion t s e e ri g u r e 81.
C oo l da* " ( " ' ' ' '
the s e c u rv e s a pply to all he a ti.ng r ate s u p to 100
- F / h r.
20'F/hr.
w e re c om pute d f o r c ooldown ra te s o( O 'T / h*r ( s tea dy s tat,),
I
i 28 60 'T /hr and 100 'T!hr.
The 20 *T/hr curve would apply to cooldown rates esp to 20'T/hrt the 60*T/hr curve would apply to rates from 20'T to 60'T/ hrs the 100*T/hr curve would apply to rates from 60'T/hr to 200'T/hr.
The Unit No. 3 heatup and cooldown curvus for.up to 5 ETPY are given in Figures 10 and 11. Urdt No. 3 carves covering 5 to lo ETPY are given in Figures 12 and 13. Corres ponding curvus for Unit No. 4 are given i
in Figures 14 through 17.
s I1 4
'Ai 4
. 5 l.'i 6
l t
,.3 li l-l l
i
_ - - _ - - -. _ -.. _ _ -.. _ _ - - ~ - - - - - - - _ - - - - - -, - - - + - - +, + _ - ~ - - - - - - - -
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f NUREG/CR 2837 !
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- _ __J
- i PNL Technical Review of
,tj Pressurized Thermal Shock issues
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Manuscript Completed: June 1E9 Date Published: July 1982 j
4 Prepared by I
L.T. Pedersen, W.J. Apley, S.H. Shn, L.J. Deffording, M.H. Morgenstem, i
P.J. Pelto. E.P. Simonen, F.A. Simonen, D.L. Stevens, T.T. Taylor Pacific Northwest Laboratory o
Richland, WA 99352 i
Presared for Divlsion of Safety Technology l
l -
Office of Nuclear Reactor Regulation t
U.S. Nuclear Regulatory Commission 1
4 Washington, D.C,20665 NRC FIN B2510 l
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ABSTRACT Pacific Hortnwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS).
Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be requircd to prevent an unsafe operating condition.
PNL revitwed these responses and the results of supporting research and concluded that none of the eight reactors would undergo vessel failure from a PTS event before several more years of operation. Operator actions, however, were often required to terminate a PTS event before it deteriorated to the point where failure could occur.
Therefore, the near-term i
(less than one year) reconwendation is to upgrade, on a site-specific basis, operational procedures, training, and control room instrumentation. Also, uniform criteria should be deveioped by NRC for use, during future licensee analyses. Finally, it was recommended that NRC uograde nondestructive inspection techniques used during vessel examinations and become more involved in the evaluation of annealing requirement s.
4r 1
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1.0 INTRODUCTION
1.1 BACKGROUND
The pressure vessel of a nuclear plant is subjected to a pressurized ther-mal shock (PTS) when an extended cooling transient to the vessel wall is acconw panied by system pressurization. Under these conditions, thermal and pressur-12ation stresses on the internal surfaces of the vessel are additive. Moreover, these stresses are in tension and tend to open cracks located at or near the internal surf aces.
Nuclear plant pressure vessels are f abricated from ferritic steelt. The i
internal surf aces of the vessels are clad with stainless steel weld to prevent metal corrosion processes. The vessels are designed to withstand Hermal heat-ing and cooling transients for the life of the plant, which is usually 40 years r
'I at 807. opersting efficiency or 32 ef fective f ull-power years (EFPY). A pres-sure vestel intended for 32 EFPY must be designed to maintain f racture tough-
?
ness of the vessel material. An adequate level of ft6cture toughness prevides assurance tnat small cracks will not propaghte in r. " brittle tranner as a result of strcsses Arty en abnormal transint such as a PTS event. Gilure in a brittle motett could feat %rv the veuei wall and lead to sesere f ailure of the pressure boundary Tr. the core area.
la contrasts a ductile type of f ailure world t,e aspectM to result, at worst, in a thro 9gh-vesbel crhck, which would leak but not result in a total loss of tne pressure bcsundary.
t r
in older nuclear plants, the pressure vessels were of ten f abricated with weld materials :oritaining relatively higr levels of cooper, pnc%phorus, and nickel.
These elements were later shown to result in greater irraciation danw age to the vessel material than had beer: initially expected, irr adiat'.on dank age caused a shift in the fracture toughness curve to higher temperatures and, therefore, increased the remote possibility of a nonductile vessel f ailure.
Evaluating the failure probability of any nuclear pressure vessel is very l
complex. The evaluation must be plant-specific to allow f or differences in material properties of the plant components, systems configuration, operating procedures, and dosimetry history. The plant control systems, component redun-dancy, operating history, and operator training and proficiency are important in determining the initiation, sequence, and timing of accident-type events and in evaluating the probability of mitigating operator actions.
Finally, the thermal-hydraulic, material properties, and f racture mechanics analyses, using 4
currently available codes, are used to determine the consequences of the events
)
being analyzed.
The following conditions must be present during a PTS event bef ore a sig-nificant nonductile failure probability would be expected:
i n
It 1.1
The nuclear plant pressure vessel must exhibit significant loss of e
fracture toughness through neutron irradiation, An overcooling transient must occur that would be of sufficient dura-o tion to cause a steep thermal gradient across the vessel wall and cooling to the low-toughness temperature range, A flaw must be present of sufficient size and be located at a criti-o cal beltline location where reduced fracture toughness and high ther.
mal stress exist.
A simultaneous high reactor coolant system pressure must be present.
e In recent years a number of incidents have occurred that involved several, but not all, of the above conditions.
The PTS issue is, therefore, being investigated in much greater detail by the NRC, the utility industry, and Nuclear Steam Supply System (NSSS) contractors.
1.2 GBJECTIVE OF,, STUJY, I
heific horthwest I.aborstury is providing technical assistance to NRC to de% lop and recommend a regulatory position that NRC should adopt before the J
longer-term PTS orogram provides genarte resolution and e.cceptance criteria.
The near-term *ce.unendations include any corrective acticg))reqJired at tht eight plants idenO fted in the August 21, 1981 NRC letter.U The recomen-dations of this report are cased on the review of information described in Sec-tion 1.3.
t L3 APPROACH i
Eight pressurized water nuclear power plants (Ft. Calhoun, H. B. Robinson 2 San Onufre 1, Maine Yankee, Oconee 1. Turkey Point 4, Calvert Cliffs 1, and Three-Mile Island 1) have been identified for specific review of PTS event scenarios.
These plants and in response to NRC requests.yhg+
SS owners groups have supplied information The following sources of information
+
were used by PNI. to recommend NRC's near-term regulatory position.
l 1.
Documentation by the licensees and owner groups to the NRC requests for information concerning the PTS issue.
2.
Participstion in reviewing current procedures, training, and operator responses to PTS events at selected plants as established by the NRC's PTS task force on procedure review.
3.
Reviews of rosearch work being performed in support of the PTS issue at NRC, national laboratories, industry, and other research institutes.
1.2 x
^
5.0 MATERI ALS PROPERTIES OF 1RRA01 ATED VESSELS Pressure vessel steels exposed to neutron irradiation experience a degra-dation in fracture resistance.
Ferritic steels have an intrinsically poor The loss of ductility with decreasing fracture resistance at low temperatures.
temperature occurs as the nil-ductility transition temperature is approached.
Below the transition temperature materials f ail by unstable, brittle fracture, whereas above that temperature materials f ail by stable, ductile fracture.
Neutron irradiation causes the nil-ductility transition reference temperature The shift can be large enough to (RTNDT) to shift to higher temperatures.
endanger the integrity of the pressure vessel if the irradiation-shifted nil-tamnarature of the vessel ductility temperature is elevated abova the seru4ca talb Of particular concern is the fracture resistance of irradiation-sensi-tive welds.
Two factors aggravate the fracture resistance of irradiated vessel welds, subjected to a PTS event, in some cases, aggravation occurs when the irradia-r tion history of t@_ reactor has resulted in sienificant elevation of the nil-tamnerature.
In other cases, aggravation occurs when PTS lowers the ductility
[
wall temperature, wnich thus lowers the fracture resistance of the vessel j
3 welds. Accurately preficting the f racture of a vessel weld recuires estinating j
the vessel neutron exuosure histories, welding pro.?.0dures, and the irradiation g
sensit kities of welds as a function of chemistry. Furth*rmore, the radial dependence of nevron *,pectrum and flux in the wall must he evahated to quan-
{
titatively determine the increasing f racture toughness through tte wsi!.
This chapter describes the effects that irradiation and mhterial charac.
teristic, have on the degraced fracture resistance of presswc esal steels.
1 Methwis used by licensees and owners groeps to predict fracture revistance aN
}*
f the uncertainties inherent in these methodt, wc evaluated.
L astly, the State i q, of knowledge is evaluated to indicate what inf ormation may Mcome hvailable in lb the future which would aid in evaluating the integrity of irradiated pressure Ut L !
vessels during a PTS event.
f t
5.1 NEUTRON 00SIMETRY
,,e Atomic displacements caused by neutron irradiation are the principal cause of degraded fracture toughness of nuclear pressure vessel steels.
The degrada-p{:
tion is directly related to the number of high-energy neutrons that penetrate
[
the steel. Traditionally, the number of neutrons having an energy greater than 1 MeV has been used to characterize the irradiation exposure.
Predicting the material properties of plant-specific reactor vessels requires an accurate E
i knowledge of neutron exposures of metallurgical test specimens and an accurate d
i knowledge of the neutron exposure of plant-specific pressure vessel components, I
Methods used to irradiate and test metallurgical specimens and to estimate M
neutron exposure of vessel components result in uncertainties that affect the 5
L h
Y:
y 5.1
r r
k predicted reliability of vessels during a PTS event.
Accurately defining the neutron irradiation environment requires knowledge of the neutron spectra, flux, and fluence, as well as the irradiation temperature.
IrrJdin un of surveillance specimens provides the most reliable data base for predicting the irradiation properties of vessel components.
Such data have the most credi-bility,l wall.because they most accurately represent the neutron environment inside a vesse The plant-specific neutron spectra and fluxes are similar for surveillance irradiations and inner-wall vessel irradiations, i
Methods used for vessel dosimetry are dependent on dosimetry analyses of surveillance capsules and on calculated neutron fluxes. Discrete Ordinate Transport (DOT) codes are used by the licensees and owners groups to map out the spatial dependence of neutron flux.
The calculated fluxes are then com-pared with measured fluxes using flux monitors inserted in surveillance cap-sules.
The DOT codas are considered to be accurate, but if wrong input values are assumed, the predicted fluxes can be inaccurate. When predicted flux compared with measured fluxes, the values can agree to within 10% to 15%.b$ pre l
The uncertainty in peak fluence values provided by the licensees and owners groups is reasonable; the values for Combustion Engineering were within 30%,
the values for Westinghouse were within 20%, and the values for Babcock & Wil-cox were approx 5etely 15%. The discrepancies in peak fluence values represut encertainty in the predicted peak fluence (E > 1 MeV) at the inner surf act of the steel venel.
Additional uncertainty can exist in the predicted vessel properties beccuse irradiatten tests and vessei wails have different neutron spectra and f i ta e s.
These differences are minimized when the properties of surveillance spe:imen are correlated to vessei properties.
The correlation is possible because the neutrco spectrum and flux of the surveillance iocation are similar to those four.o inside the vessel wall. When proje ting properties through the thickness of the vessel wall, the spEtrum arrd flux are degrt.ded.
Ttu spec-trum is shif ted toward a lower average energy with many neutrons below 1 MeV
- )
contributing to irradiation damage.
i:
h 1:i To account for these lower energy neutrons, it has been recommended that j'
9 displacements per atom (DPA) be used as a measure of irradiation exposure.
The damage based on DPA is greater through the wall than would be predicted based on the E > 1 MeV assumption. Differences between the two exposure c ria as o
a function of distance through a vessel wall are given in Table 5.1.
g F
As radial distance increases, damage rates decrease.
The lower damage c
rates may provide a greater opportunity for self annealing during irradiation.
I Hence, damage accumulates more slowly per OPA for positions deep in a vessel wall.
This suggests a lesser damage in deep regions than would be expected if rate effects on damage efficiency were neglected when predicting radiation damage through a vessel wall. The effect of the damage rate efficiency can be estimated by comparing damage rates with thermal annealing rates.
Combinations b
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5.2 l
l t
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i2" s.3.h j,3' j fj NUCLE AR Rf GUL A10HY COrdtalSSION i
3
.uf...mos on. o. c. to a
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December 3, 1982 1
f.
P.EMORANDUM FOR:
Commissioner Gilinsky Commissioner Ahearne r< -
g,#
TROM:
Demetrios t..
Basdekas l
l Instrumentation & Control Branch N"lW" I
Division of facility Operations Of fice of Nuclear Regulatory Research c '.
SUBJECT:
staff REPORT ON PRESSURIZED THERMAL SHOCK, SECY-82-465. NOVEMBER 23, 1982.
lt Earlier this week I discussed with you a number of, what I considered to be, l,.
significant points on the subject staff report.
Unfortunately, access to the report was denied to me until the afternoon of November 29, 1982.
Based on my limited review of this final version, I have prepared the, following summary of the points I discussed with you including a few additional ones.
1.
The probability of a PTS caused vessel rur,ture and core-melt is not quantiheM e wi th
- car +ainty which can form the primarv bWis for decision.gkinc on this matteE I belieQnQWITEco ort the qua?iTITit'ive risk es:T5htes by the staff is unsiirranuCJ also believ~e that thEGYf shiuld provide a complife~Fthorough response to the two basic questions:
c (a) What is the uncertaiaty in the estimate of the probability of PTS-caused catastrophic vessel rupture and core writ'l (b) What is the confidence level in that uc.certhinty f HW was i,t dtrived?
(See pp. 3-9, H-26. Sec. H 4.1 of Staf f report) 2.
The P.:ncho Seco cycnt was considered by the Staff to be the most severe, and so stated on p. H-26 first full paragraph of the main report.
The Crystal River event as described in Section 2.2.6 and as shown on Figure 2-13 of the main report appears to be more severe than the Rahcho Seco event.
I belicyc that the Staff should provide a definition of severity and answers to the following questions:
(a) is the Rancho Seco or the Crystal River event the most severe?
(b) How confident are we on the time history data of temperature and pressure that have been provided by the licensees?
(c)
If D.ancho Seco is not the most severe event, how does this affect the analyses performed on the assumption th,at it was?
t
7 l{
Commissioner IhearnE
~
i 3.
In-situ annealing capability of PWR vessels has not been demonstrated l4 and there is considerable doubt that it will be available for a long time, if ever, during the lifetime of.most PWRs which have exceeded i
i or are expected to e.xceed 200'T of their vessel RT There are also j
plants that are facing similar difficulties with regard to an iiDT.
1 t
acceptable limit of the upper shelf toughness.
I believe that the
'y following question should be answered in some detail:
(a) Which plants do not meet existing regulations? (ite., Appendix G. IV.
1
(
A2a, B, and C relatin annealing capability)g to upper shelf toughness and in-situ 4.
The staff acknowledges the importance of Instrumentation and Control Systems malfunctions in PTS (See second full paragraph beginning on p.1 of SECY 82-465), but it has not asked the utilities to supply design information on these systems and their electrical power supplies D
in its letters of August 1981 and since then (second full paragraph on page 3 of SECY-32-465).
The following questions have been asked by the Commissioners before, in one form or another, but no definitive
],
answer has been provided, to my knowledge.
i.-
(a)' What is 'the reason for this inconsistency between the stated i
l.
importance of instrumentation and control systems (p.1) and stated tettens
'p.
3)?
l,l -
(t)
If we co not have a timely and technically sour.d resolution of USI A-47, Shfety lep',1 cations of Ontrol Systems, how can you
{
expect to resolve A-49. P % ?
Further;no e.
(c) Without Jnign information on the plants we have chosen to review ut. der both USIs A-47 and A-49 (Oconee41, Calvert Cliffs-1, H. P. Robinson-2 )
how can we justify' the large expendituies of our RES and tiRR programs which deal with 15C systems initiated transients of importance to PTS or any other safety is. sue?
i 5.
The proposed screening criteria are 270'F for longitudinal and 300'F for circumferential welds in the. RPVi.
The selection of the screening i
criteria method is based on eight events taken on a cumulative manner of all PWR experience.
This leaves out probability components Associated with (a) substantial operational experience involving event sequences which terminated early enough or in some other benign way, which might, with some probability, have continued on to produce a more severe challenge to the RPV and (b) essentially those potential events and their associated sequences, which have not occurred yet, but which may, with some probability, occur in the, future, causing a severe challenge to the RPV.
These are important conside, rations in estimating probabilities of event sequences that i
c Summary of Weld Properties and RTNDT Predictions TABL E 5.2.
- F n/cm Date Cu, %
Ni.'%
Mean + 2e Plant 1
1.10 x 10 '
9/30/81 0.30 0.57 265 rtey. Pt. 4 Circum.
+20 18 Fort Calhoun Long
-20 6.48 x 10 12/31/81 0.35 0.99 268 2 410 19 San Onofre 1 Long
-20 2.75 x 10 10/31/81 0.35 0.20 278 7 860A 18 Calvert Cliffs 1 Long
-20 7.05 x 10 12/31/81 0.30 0.99 267 2-203 18 t4aine Yankee Long
-20 4.73 x 10 12/31/81 0.36 0.99 251 2-203 1I Robinson 2 tong
-20 1.30 x 10 9/30/81? 0.34 0.20 218 2-273 II Cire
-20 1.24 x 10 (assumad)0.34 0.!0 253 11-273 18 Oconee 1 tong
+2C
?.27x10 10/01/81 0.31 0.55 183 3a.1430 0.27 (a) RTNDT (HE06)
RT0,(. (,8 + A70*Cu + 350*Cu*f41)* (.--[.g) 9
'10
.a g
e 5.3 1RRA01AT10N PROPERTl!$
The shif t in the nil-ductility temperature due to neutron irradiation of The issue for the PTS evaluation is to pressure vessel steels is well known.as aerurately as poss1 Die Tor spec 1TTc vesiel i
.1ht_.lf.r.ad13t 4 ^" c h4 f t quanti,fj_Because specimens cannot be extracted from the irrecTatea vesseis, it is necesUtty_.lo prgioct irradiation properties f rom irradiations of metallurg-we]A
'u ical test soedmem. The irradiation environment and materials used fo
[
met W gical specimen irradiations must approximate, as much as possible, the Furthermore, irradiation materials and environment of the pressure vessel.
i tests must project the properties at some future date--in particular, to end t
a of itfe or 32 EFPY.
The irradiation tests that were used to estabitsh Regulatory Guide 1.99, Rev. I were performed primarily in test reactors at enhanced fluxes and in neutron spectra having average energies larger than those typical for pressure The rapid fluxes meant that fluences in end-of-life reactor vessels vessels.
5.7
+.
,N p
L*
w........ v., s...........
may cause a PTS of certain severity in the future.
Hence, plant-specific analyses are needed to estimate a meaningful number of e
I believe we should consider probability and RT ni screening value.
a criterion for eac5 vendor design and ultimately a limit for each N
l*
(See Co m.ent No. 6, below).
These limitations are acknowledged M
plant.
by the staff but their significance apparently is somewhat elusive
'M.
'[q'-
when it comes to formulating the conclusions and reconmendations for the screening criteria.
3..
An important question that should not escape serious consideration c
is-(a) How do we reconcile this selection of screening criteria with the fact that a Small Break LOCA is capable of cooling down the vessel to about 125'F within about 30 minutes with a subsequent isolation and.epressurization to full design i.
pressure?
The summary of operational experience given in Section 2.3, Figure Il
!. c.
6.
2-14. Figure 41, and elsewhere in the Staff report, provides a l i '~
lumping of the o;erational experience for reactors designed by all thr,ee vendors.
This results in a " smearing" or " averaging out" of the operational data associated with reactors of individual vendors, i
l I believe that a meaningful PTS assessment may be performed on a plant-specific basis only, and with substantial limitations on a 1-vender. ger.eric basis, but, I believe, with almost nil utility for i
an t.ll-vend:r-ge.9ric bas is, hence, the selection of the screening criteria di<,cussoc in Cha ter 4.0 is. Msed cr. *very weak <jrcunds.
.L (Sce Comment No. 5, above I believe that the following question should be answered by the Staff:
V Would you explain how, in your judgment, the lumping together (a) of operational experience from plants supplied by all three NSSS vendors (B&W, W, and CE) gives you a realistic and applicable l,[
data base for all of them, when you consider the fact that the l;
dominant contribution comes from B&W plants PTS precursor i,-
events?
What is the combined effect on your selection of screening (b) criteria when you take into account the consideration of the points discussed in Comments 5 and 67 1
e y
+
I I
vi c o..
A flux redretion by a f actor of 2 4.5 gives aMing[imated RT N0f' 1his is well within the uncertainty band of the es 7.
flux (a) Does the Staff think that a more meaningful and o
F l nts that reduction would be one by a factor of 10-30 for p a flux reduction is needed?
ts are not In service inspection techniques and frequency requireme
.,y.
i l
very effective in producing useful and t me y 8.
g',,
equest the related analyses.
I believe that the Commission may find it appropriate to r Staff me:nbers'
/
Staff to address this issue in some detail (including ii stated by the opinions, which may be in variance with the pos t on Staff during the December 1,1982 briefing).
h September 13, 1982 The enclosed memorandumII) contains my comments on t eha draft of the subject Staff report. A number of them the most impor propper consideration in the final report. However, have remained unresolved.
Comnission consider the My only recomendation at this time is that the i dinal and circum-following interim ' screening criteria for both lo be governing:
of 150V RTNOT For Babcock and Wilcox Plants:
RTNDT of 200'F For Westinghouse and Co.nbustion Engineering Plants:
i listic, timely, and These screening criteria would provide for inore rea
\\
prudent resolution of this issue.
i participation in Dr. Okrent's recommendation (2)for the Commi "establishir.;(the criterla 'o be used on this issue j,
h tsr.phasis addcd) is vnry appropriat.
most ot' my canents unceQainty" nt
!. appreciate the opportunity to have discussed with youque made above, and I will be pleased to answer any iteria.
3 as well as my recommendations for interim screentnp cr s
to your ding requests By copy of this memorandum I am confirming my pento me colleagues on the Commission for the opportunity
- issue, them. individually my views on this important b n -e.h A l-8as.d d m',,
Ocmetrios L. BasdekasIns Division of facility Operations
/
l l
I.
'tonsuiss Ener mesieiv I
Enclosures /Re ferences :
Memorandum from D. L. Basdekas to P. S. Shewmon, ACRS, October 6,1982.
1 1.
(Enclosure) 14, 1982 -
Letter from P. S. Shewmon, ACRS to Chairman Palladino October 2.
Additional Comments by ACRS Member David Okrent (Reference) i t
t cc:
Chairman Palladino Commissioner Roberts l
Commissioner Asselstine W. Dircks, EDD V. Stello, CRGR T. Marley, CRGR e
H. Denton, ttRR F. Schroeder, fiRR R. Minogue,. RES D.<Ross, RES i
K. Galler, RES E. Wenzinger, RES Y
Il
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e P
g D
a n
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l-l l.
7 6
LEHIGH UNIVERSITY j
Institute of Fracture and Solid Mechanics
~
Packard Lab. Bldg. #19 BETHLEHEM. PENNSYLVANIA 180ls Te:ex No. Lehigh Univ. UD 710-6701066 we,
I j i. - _. __
~
+
O. C. Sih Director October 10, 1985 i
Atte me/ Martin H Hodder 1131 N.I. 86th Street
- Miami, lorida 33138 RE:
Turkey Point Nuclear Power Plant Unit No. 4:
Reactor Vessel Embrittlement at: Surveillance Program
Dear *tterney Hodder:
Ir response to your letter dated August 29, 1985 and the above referenced i
subje:t. matter. I have read tr.e package of docunents en the RPV smbrittle: rent y
prograr at Turkey Point Unit No. 4.
A nu.tber of supporting aegunents with ref-erence to the calculation of ART are questionable, if not invalid from the l
NDT scientift: view peint In what follows, the SWRI report and the TFL letter shall bereferredtoas[1)*and[2)**,respectively.
(* ;
SWRLPydictionE1],
i Based on the RPV material surveillance methodology, SWRI [1] estimated the sh Ut in RT for Turkey Point Unit No. 4 The results pertaining to wall NDT locati:P 1/4T based on the data of Capsule T in terms of EFPY are surrnarized graphi:111y on the sheet attacned to this letter.
The shift in RT is found NDT to be 1:oroximately 324'F at 8 EFPY.
This is beyond the NRC screening value of 300'F.
- E. B. Norris, " Reactor Vesse! Material Surveillance Program for Turkey Point Unit Nc. 4:
Analy is of Capsule T", Southwest Research Institute Technical Re-port Nc. 02-4221, June 1976.
..Letter, Uhrig. FPL, to Eienhut, "Re:
Turkey Point Unit 4, Docket Nos. 50-251, PTS to Reactor Pressure Vessels", January 21, 1982.
E.EEj048
$5 I' j f
. No.i g _,m_'
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w-2-
(2) FPL Resoonse [2]
t f
With reference to the material in Docket No. 50-251 on PTS of RPV as stated in [2), a lower ART value of 211'F was obtained for Unit No. 4 This NDT result, however, was obtained by application of the surveillance data taken from Turkey Point Unit No. 3.
The justification was that the metallurgical properties of the beltline welds of the Turkey Points Units No. 3 and No. 4 are the same and that data on Unit No. 4 are not sufficient.
(3) Comments The rate at which the beltline weld material deteriorates and/or em-brittles depends on the combined effects of irradiation and pressurized thennal shock.
It is plant-specific in the sense that the influence differs inherently from one unit to another.
In other words, the metallurgical properties alone cannot ce: ermine the damage behavior of the welds. The loading histom/ plays a major role.
Unless the rates of irradiation, fluctuations in thermal gradients and time variation in pressure are exactly the same for both Units No. 3 and No. 4, one is not justified to assume that data collected in Unit No. 3 could be applied to predict the behavior of Unit No. 4.
Hence, conclusions drawn on ART for Unit No. 4 based on the data of Unit No. 3 cannot be considered valid.
NDT I will not delve into the other details concerning the actual calculation af ART as they are beyond the scope of our immediate concarn.
NDT Very sincerely yours, 5
,d !!
! -j
'3 4
'GeorgeC.hdh Professor of Mechanics GCS:bd Enclosure
p.,
p.
Data-Reproduced from Table on Page 3 at Uall Location 1/4T, Report by E. B. Norris, " Reactor Vessel Material Surveillance z'
Program for Turkey Point Unit No. 4:
Analysis of Capsule T",
Southwest Research Institute Technical Report No. 02-4221 June 1976, t
500 -
450 400 27 2,
EE
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$i 350 f
.li
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cf
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324'F 300
- NE Sr.reening Criterior.
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250 I
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I 200 I
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5 8
10 15 20 25 30 Effective Full Power Year (EFPY)
6 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM i
CAPSULE S - TURKEY POINT UNIT NO,3 j
1 CAPSULE S - TURKEY POINT UNIT NO,4 "i
1 i
i FINAL REPOR i 1
SwAl Preject No.C.5131 SwRI Prc'ee: No. 02 C30 L
)
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0.
- 8 j-
{
7 F!crica Pewer & tight Ccm;:ar:y I
F, O. Ecz 3100 I
Miama, Flen:a 331C1 f
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- R N.%s-S=j s o u T H w i s
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1.A s 4 %
- 0 %
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- 0. 3
- 0 s
a reccter vossol.( 1) Tho projected f aat neutron exposures resulting.
the analyses of the second surveillance capsule (S) fica es:h uni:
h ar, a
in 300d agree ent With th se re;Cr:ed earlier.(
- 3)
A13o, since the j
S capsules did not contain spect= ens representing :he centro'. ling (weld
= eta.1) beltline material, there is no basis for revising the prejected
[
(I values of RTg; used to develop the curren: set cf heatup and cooldevn t
1:=1: curves.
- el I
i F.
ca-sule p.azeval schedule
.g li A third :apsule is seneduled f r re= v11 fr:m ea:h reac::t vessel f
at:er 10 calendar years of c; era:1:n.
3ased :n the ;as: :;erating histo-ries of :he Turkey Poin nuclear ;cvar ;'an:s,10 calendar years :! :; era-E
- 1:= should c:rres; nd to appr:xt:ately 7 ITPY cf operatirn.
I:fsye'in-i
.. v. 4..., w ~
sr...3 W ma:s!'specizer.s.
zer.ded that. Capsule Y,a Typ e !! cap sul...
e. ;:J.,: a'j:
- q
=
de ra= ved fr:= each vessel'it th'a: ti=e.
The pra,iected ias neutren f2u-l ent e f er the *.' =ap sules a.*:er 7 IFM* is 1.1 x 10'9 :='2 (I
- 1 Me'.'), a;-
l9 pt:ximately tvi:e :he fluence received by the T cap sules. (la l5) ' he da:a U
e l
.s
- .:atted fic: the V =apsules should pt: vide :he internati:n ne':essary,;a revise :he hes:up and c e.'d:.n lisi:a:::ns' f:r etera:1:r. tey:nd 1: IT7T
- ! : eratien.
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4 UNITED STATES
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NUCLE AR REGULATORY COMMISSION 1
p wateantecTO8v. D. C. e9666 3
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April 22,1985 D'*k't**8;!$!!!
?PG3fPnn nrP]
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1185 q MAY J SGOL5U U LW Mr. J. W. W1111tms, Jr., Vice President 14ewmati & Holtzinger Nuclear Energy Department Fiorica Power and Light C:mpany
/
8:st Office Eex 14000 Junc Beach, Ficrica 334:3
- ea-Mr.. Willises
e ::-rissi:n has issue: the enc 1: sed Aree: ment No.112 te Facility
- e-atin; Licease Nc. *,P:-31 and Arencment No.106 to Facility Operating
' ice se NO. OPR d'. f:r t*e Turkey Pcint Plant Units Nos. 3 and 4,
- es:ectivety. The arencNnts consist of changes to th( Technicil 5:eci'ications it, res:ense to your e:rlication transmitted by letters l
- stec Feerua y 5,1925 a.: Mar:n 6, HSS.
i
- hese arendments revise tne ie:hnical Specifications te pr: vide consistency in icentificatien of the surveillance specteen :npsules in the Technical l
L 50ecifications anc tne actaal surveillance specimen capsules. The surveillance s;ecimen enmination sc"edule is also modified to provide bettee in" creation in ac:ordance with the current regulations. The cr00csed changes c:moine tne existing Reactor Materials Surveillance Program into a single integrated program whien confoms to the requirements of 10 CFR 50, Accendices G and H.
We have discussed concerns and actions necessary regarding future core designs and in cavity dosimetry in Section III Of cur Safety Evaluation provided in support of the amendeents.
Section II.C of 10 CFR SC Accencix H, which was revised on July 26, 1983, Oe-its an integrated surveillance program provided it meets the criteria s:ecified and is a::revec by the Director, Office of Nuclear Reactor Regulation. We have incicated in our Safety Evaluation that the integrated surveillance :rogram for the Turkey Point Plant pemitted by the enclosed amenements meet the criteria specified in 10 CFR 50, Appendix H !!.C.
The Directer, Office of Nuclear Reactor Regulation, has approved the enclosed amencrents which authori:e an integrated surveillance program at the Turkey Point Plant in accordance with the recuirements of 10 CFR 50, Appendix H II.C.
l iXATBff E
~
(.b d
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i Mr. Williams April 22, 1985 A copy of the related Safety Evaluation is encle' sed. A Notice of Issuance will be included in the Comission's next regular month 1/
t Federal Register, notice, j
1 Sincerely, k
I-Daniel G. Mcdonald, Jr., Project Manager Operating P.eactors Branch #1 Division of Licensing
/
j Enciesures:
1.
Menement No.112 to OEE-31 2<
Mencmen; Nc.106 to 0E4 41 3.
Safety Evaluation cc: */tnclosures See next page' t
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UNITE D cT ATE S
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NUCLE AR MEGULATORY COMMIS$10N
$ ' kg. r %
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WASHINC T ok. o. C. atss5
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.S. AFETY EV/.tVATIO'l BY THE OFFICE OF KL'C'.!AR REACTOR FEL ATED TO At'.ENDPE!4T N0.112 i0 FAtlLITY CPERATING LICEt4SE f:0. Ci
.%D p!TREtif !:0.106 TO FACILITY OFERAT!!$ t1 CENSE R0. OPR 41 FLORICA F0WER A O LIGHT COMPA1E i
TUF. KEY PO!!!T LIIT t405. 3 At:0 4
- CCKET t.CS. 50 2E0 at:0 50 251 l
- . _:~ u: cue *i:n r bru:ry 8,
- a l e '.t e r f r :.. J. W.
111ams, Jr. t: ?
G. Sisenhut, :atec c
lii5, F'.:ri:a b cc !. t'; *, C:mpany repuetted that the Surkey Point Units Te:n.ical !:trifications be aten:ec to cc..bir.e the react:r 4.
3 er:
vessel raterial sursetilam:e program for these units irite a sir.gle inte-
- -ated surveillance pr
- ; an.
Accitional information concerning the pro-p;sec change was pr:vi:e: ty thre licensee in a 'er.-tter f t:m J. W. eli1116-s, Jr.
to 5. A. Varga dater.' Mae:h E.195E.
I A revised Appencix H.10 C8R 50 was published in the Feceral Ra;ister en May 27, 1933 anc became effective on July 26, 1983.
Section I!.C of the revised Appencix H permits an integrated surveillance pr: gram provicec it f
This j
is a;prevec by the Dirc:ter, Office of flu: lear Reactor Regulation.
se: tion of A;;endix H icer.tifies the criteria to be usec in ovaluatin; un l
integratec surveillance ;r:; ram.
The criteria are:
1.
There must be su stantial advantages to be gained, such as recuced t
pes.cr cutages or reduced perscnnel exp;sure to raciation, as a cir:::
result of not recuiring surveillance cassults in all reactars in tne set.
l I
i 2.
The design and cperating features of the reactors in the set must ec sufficiently similar to permit ac:vrate c mparisens of the predicted amount of radiatien camage as a functien of total pcaer outp.*t.
1 3.
There must be an acepuate cosir.:etry proger.m for et:n res: or i
Qf.
x:ry;.
i n
.g.
There must be a contingency plan to assure that the surveillance 4..
program for each reactor will not be jeopardized by operation at
]T reduced power level or by an extended outage of another reactor from which data are expec)ed.
5, No reduction in the requirements for number of materials to be irradiated, specimen type, or number of specimens per reactor is f
permitted, but the imount of testing may be reduced if the initial results agree vtith prtdictions.
,i 5.
There must Oc ace:uate arrangement for cata sharing between plants.
E
- . Evalust en Each. unit at Turkey P: int began :ommercial operation.<ith 8 surveillance capsules in each react:r vessel.
Ten capsules contained forging material and six capsules contained veeld metal, forging, and heat affected zone (HAZ) materials. To date, two capsules containing furging material and two capsules containing weld metal, forging, and HAZ materials were irradiatad, removed from the. vessel, and tested.
The test results from the surveillanca meterial indicate that the weld metal will sustain the most irradiation
'Since, based on the initial test, the weld metal is more l
damage.
susceptible to irradiation damage than the forging material, the licensee has proposed to retain the capsules with forging material as standby li specimens in the reactor vessel and test only those capsules with weld metal, forging, and HAZ materials.
Since fewer capsules will be withdrawn than originally anticipated, the radiation exposure (ALARA) to plant personnel should be reduced.
7, L
the licensee's F5AR Volume 2 indicates that the materials and desi ns for 0
L the core, thermal shield, core bartc1 and vessel are the same for each l-unit at Tur, key Point.
Since the neutron energy spectrum is a function of geometry, materials, and core loading, the relative neutron spectrum for l.
both reactors should be equivalent for equivalent core loadings.
The l
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6 licensee indicates that fuel management and cycle lengths for both units have.been similar. Thus neutron spectra profiles at the peak fluence l
locations should be equivalent.
The neutron fluence, which is used to predict radiation damage, is calcu-lated using PCQT power distribution data, and computer codes SORREL and
.dCT4.3.
As built timensions and individual material properties 4re incorporated into the CCT 4.3 mocels.
Hence, using these codes, the li:ensee.ill te atie t: preciet raciation damage as a function of power cutput f or each unit.
Eacn vessel has b:tn in-:sosule and in-cavity dosimetry, which will be usec to verify the neute:n spectra and the codes that were usec to predict '
neutron fivence as a function of power output.
Sir'e each plant has its own capsules anc octn plants are capaele of independently predicting and monitoring raciation camage as a function of power output, the p ogram will not be significantly jeoparci:ed b/ operation at reduced power levels or by an extenced outage of either plant.
. Based on the intial test, the limiting material fur each unit is weld material, which is identified as SA 1101. This material is in each capsule that will be irradiated and tested.
Capsules that have been deleted from surveillance testing do not contain the limiting material and will be retained as standby specimens in the reactor vessel.
Since the amount of limiting material in the surveillance program has not chnaged, the numoer of useful surveillance specimens available for testing has not changed.
Both units have common management and the surveillance program will be Therefore, there should be managed by their Nuclear Energy Department.
aceounte data sharing.
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We have concluded based on the details in Section !! of this Safety 1.
Evaluation, that the integrated surveillance program meets the evaluation i
L criteria specifiec in 10 CFR 50, Appendix H II.C.
If future core designs
~
aresignificantlydifferentthanthosedocumentedbythe)icensee,the licensee must' explain the effect that the changes have on neutron J
irradiation' damage' ana the surveillance capsule withdrawal schedute, 4
In :avity d:simetry testing should continue in orter to reduce Po-2.
If these test results jettet vn:ertaint:es in neutron fluence.
pr;vi:e an ef f ect'.e method of monitoring vessei neutron. fluence, t..e in co.ity dosimetry should be incorporated into the integrated sarveillance program.
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-5 IV. Environmental Consideratien These amendments involve changes in the installation or use of the facilities components located within the restricted areas as defined in 10 CFR 20 and in surveillance requirements. The staff has determined that these amencrents involve no significant increase'in the' amounts, WId no significant change in the types, of any effluents that may be released l
offsite and that there is no significant increase in individual or cumulative occupational racistion ex;csure.
The Commission has previously issued a prootsed finding that these amendments involve no significant hazards consiceration and there has been no public comment on such finding.
Accordingly, these amencrents meet the eligibility criteria for categerical
. exclusion set forth in 10 CFR Sec 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment yeed be prepared in connection with the issuance of these amendments.
V. Conclusion We have concluded, based on the censiderations discussed above, that:
i L
(1)' there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, p
and (2) such activities will be conducted in compliance with.the l
Commission's regulations and the issuance of these amendments will not be inimical to the common de#ense and security or to the health and safety o# the public.
Datec: April 22, 1985 Principal Contributors:
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UNITE 9 STATES 9I
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,r bruary 27, 1985 e
Docket Nos. 50-250.
and 50-251 t
Mr. J. W. Williams, Jr., Vice President e
' Nuclear Energy Department Florida Power and~ Light Company Post.0ffice Box 14000
~ Juno Beach, Florida 33408
Dear Mr. :
Williams:
Rcference:
TAC Nos. 54428 and 55035-
SUBJECT:
NEAR TERM FLUk REDUCTION PROGRAM - TURKEY POINT PLANT UNITS 3'& 4
f By letters dated March'1, 1984,' April.2, 1984, June 4, 1984 and August 22 P
1984, you'provided the integral neutron source data we requested in our letters of November'17, 1983 and July 26, 1984 We have evaluated the data
!g.
to verify the near term flux reduction resulting from your Pressurized Thennal~ Shock (PTS) program for the Turkey Point Plant.
The results of our ' initial Safety Evaluation (SE) are provided in Enclosure > 1 to this. letter.
In reviewing your near tenn flux reduction i
program, we assessed the perfunnance of the part-length burnable absorber
')
L y-assemblies' designed explicitly for flux reduction to the pressure vessel circumferential welds and concluded that the flux reduction factor is. 2.6.
~~.,.
j'
-This conclusion was based on independent audit calculations performed by our-technical consultants at Brookhaven National Laboratory.
However,'our initial evaluation did not taire into account the revised value
\\
l of the recuired fast neutron fluence for Turkey Point Plant. Unit 1 3 &nd 4,
(
to _ reach the PI5 screening criterion.
The revised value is based on7 he l
e details proviced in our SE relating to Reactor Vessel Materials Data for the
/
Turkey Point Vessels which was provided to you in our letter dated April 26,
- 1984, 3
a,
l The results of our supplemental SE, provided in Enclosure 2, indicates that j
the combination of the new fluence value and the present loading flux reduction will allow both plants to operate for 32 Effective Full Power Years (EFPY) without reaching the PTS screening criterion.
The 32 EFPY is l
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49 Mr. J. W. Williams, Jr. Februa ry 27, 1985 F
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equivalent to the 40 year licensed life considering a conservative capacity i
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factor of 80%.
This conclusion is based on the current low leakage loading i
?-
factor. This completes our review of your near term flux reduction program.
7 Sincerely.
)
./
p-
' Steven A.
rga, Chief J
Operating Reactors Branch #1 Division of Licensing 6-
Enclosures:
As stated cc w/ enclosures:
See next page.
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J. W. Williams, Jr.
Turkey Point Plants
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Florida Power and Light Company Units 3 and 4 1
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cc:
HaroldF;Reis.Equire Administrator-
' Newman and Holtzinger, P.C.
Department of Environmental 1615 L Street, N.W.
Regulation Washington DC 20036 Power Plant Siting Section State of Florida Mr. Jack Shreve 2600 Blair Stone Road Office of the.Public Counsel Tallahassee, Florida 32301 Room 4, Holland Building Tallahassee, Florida 32304 James P.' 0'Reilly Regional Administrator, Region II~
Norman A. Coll, Esquire U.S Nuclear Regulatory Comission i
Steel Hector and Davis Suite 2900 4000 Southeast Financial 101 Marietta Street Center Atlanta, GA 30303 j
Miami, Florida 33131-2398 i
Martin H. Hodder Esquire
.1131 N.E.-86th Street Mr. Ken N. Harris, Vice President Miami, Florida 33138 Turkey Point Nuclear Plant l
Florida Power and Light Company Joette Lorion P.O. Box 029100 7269 SW 54 Avenue
'i Miami, Florida 33102 Miami, Florida 33143 i
Mr. M. R. Stierheim Mr. Chris J. Baker, Plant Manager County Manager of Metropolitan Turkey Point Nuclear Plant Dade County Florida Power and Light Company 1
Miami, Florida 33130 P.O. Box 029100 Miami, Florida 33102 Resident Inspector Turkey Point Nuclear Generating Station Attorney General U.S. Nuclear Regulatory Comission Department of Legal Affairs Post Office Box 57-1185 The Capitol Miami, Florida 33257-1185
' Tallahassee, Florida 32304 I
Regional Radiation Representative Mr. Allan Schubert, Manager EPA Region IV Public Health Physicist 345 Courtland Street, N.W.
Department of Health and Atlanta, GA 30308 Rehabilitative Services i
1323 Winewood Blvd.
Intergovernmental Coordination Tallahassee, Florida 32301 and Review Office of Planning & Budget Executive Office of the Governor The Caoitol Building i
Tallahassee, Florida 32301 l
1
ENCLOSURE 2 t -
.D l.i f'V O
TURKEY POINT UNITS 3 AND 4. EVALUATION OF THE
-f FLUX REDUCTION FACTOR USING PART-LENGTH BURNABLE ABSORBER ASSEMBLIES TO MEET THE NRC t
PRESSURIZED THERMAL SHOCK CRITERIA Introduction' The staff identified several plants in need of flux reduction in order for them to be able to operate for 32 Effective Full Power Years'(EFPY) without violating the NRC Pressurized Thermal Shock'(PTS) screening criteria. (1, 2),
for Turkey Point - 3 and 4 the staff estimated (for the end of 1982) that the
' required flux reduction needed for either unit to operate for 40 calendar years '
(at a load factor of.8).was 4.5.
Florida Power and Light (FP&L) the licensee has implemented a flecace reduction program consisting of low leakage fuel load-('-
ing patterns coupled with part-length burnable absorbers, located so as to re-duce the neutron flux to the pressure vessel circumferential weld from high importance core locations.
l L
i Based on power and exposure distributions supplied by FP&L (3-7), the Core Performance Branch performed an evaluation of the fluxes (and fluences) associate,d with the first nine cycles of operation of Unit 4 and the first 10 cycles of operation of Unit 3.
The review and evaluation included independent audit calculations carried out by staff consultants at BNL.
Evaluation Fast neutron flux (E > 1.0 MeV) calculations at the inner surface of the Pressure Vessel (PV) on the lower core belt circumferential weld were based on the flux synthesis methodology (8).
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This approach consists of the following steps:
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Detemine the contributions to the flux above 1.0 MeV near O' (the: peak azimuthal flux location) on the inner surface of the PV from individual assemblies in the reactor core based on calculations in (r.e-) geometry.
)
l b.
Detemine the contributions to the fast flux at the lower-to-intemediate shell circumferential weld from discrete 12 in, high axial segments for the two outemost rows of assemblies basedoncalcualtionsin(r,z) geometry, l
[
c.
Combine the results from (1) and (2) with the three-dimensional i.
core power (neutron source) distributions to obtain the desired flux and fluence values.
l h
The same approach was also used for H. B. Robinson and the (r.e-) geometrical i
i L
results have been used here as well. These results were generated with the 3
[
DOT-3.5 (9) discrete ordinates transport code in the fixed-source mode with an
)
$# angular. approximation.
Region dependent, 16 neutron group cross sections 8 3
~
were based on the DLC-37/EPR (ENDF/B-IV) library (10). HBR-2 has virtually idehtical' core / internals / vessel dimensions and materials to those of the Turkey Point units; therefore, the only modification to the HBR-2 results was a slight
- j increase in the flux values to account for the higher temperature of the bypass s.
water for the Turkey Point units. The results of these calculations provided
)
the flux above 1.0 MeV at the inner surface of the PV near the core major axis l
due to unit sources located in assemblies 6, 7, 8, 13, 14, 15, 19, 20 and 24, Figure 1.
Calculations were also performed in the (r,z) geometry with the reactor axial configuration as shown in Figure 2.
This configuration was modelle' with 91 d
axial and 78 radial intervals with the DOT-4.3 (11) discrete ordinates transport code.
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-3.'
The 16-group, P cross sections were the same as those used for thekr.e-)
3 j
calculations. Note that a single set of cross sections was used for the j
core, i.e. exially zoned burnable absorbers were not accounted for. : Fixed source calculations were perfomed in the adjoint mode with an 58 symetric quadrature. The fixed source was located at the inner surface of the vessel i
at the elevation of the limiting circumferential weld (Figure 2) and the
{
importance of 12 in, high axial segments in the first and second outermost j
rows of assemblies to the fast flux at the weld were determinea. Finally the (r v) and (r,2) geometry results were combined with the core power distributions to obtain the flux above 1.0 MeV at the limiting circumferential weld near the l
[
core major axis. A further refinement was included 1.e. an' exposure correction I
based on the analysis of Reference 12.
b Power and exposure distribution data were provided by FP&L for the detemination of the sources'to be used in the evaluation of present and projected EOL fluences.
While the information that was provided was relatively complete for Unit-3, not
(
all the necessary assembly exposure data were available for all cycles of Unit 4.
i Consequently, reasonable estimates were made for the average exposure associated with the peripheral assemblies for cycles for which this data had not been provided.
The only other area where approximations for the source were made for both units was; related to the axial power distributions since data were not provided for all assemblies required in the flux synthesis scheme.
Results for the fast flux at the limiting circumferential weld near the core major axis are presented in Table 1 for Turkey Point Units 3 and 4.
Results are for Cycles 1-7 (based on single exposure weighted source and exposure dis-tributions) and for Cycle 8, and 9, and for Unit-4, Cycle 10, explicitly.
Two sets of results are given for each cycle, one assuming a unifom nominal exposure of 6,000 MWD /MTU for all assemblies, and one where the assembly-wise neutron sources were corrected for the specific exposures associated with$ch assembly.
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$ The results in' Table 1 account for the neglect of pin-w based on a generic effects'on'the (r,+) DOT calculation by an approximate facto
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dtudy of.this effect (12).:
factors for cycles greater than M
P i
texposure correction, and' fast flux reduct on l
given.
Cycle 7, relative to the results for the averaged Cycle 1-7, are a
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fter each cycle and at
-The' associated estimates for the accumulated fluence aThese val Table 2.
.EOL (assumed to be 32 (UPY ) are given in lts indicate 9
on.the exposure corrected fast flux values of Table 1..
The resu be achieved at the thatasignificantreductioninthefastflux('62%)can l loading pattern critical weld. by a combination of an " extreme" low leakage fue(inass
' coupled with appropriately located part-length absorbe o
15'ofFigure1).
the average Cycle 1-7 El
~( -50%) relative to the value obtained by assuming that L,
N power distribution is applicable through life.
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If the A reduction of the fast flux by 62% is equivalent to a factor p
- e t 9 in Unit flux reduction which was implemented for Cycle B in Unit 3 at.
I i
i n in 1989.
- ' '4, were maintained both units would reach the screening cr te
- l iJ
_(assuming an 801 load factor) (13).
d ction factors were presentation to the staff, progressively higher flux re uA fl jl
~
h.
However, our estimate of the planned for both units.
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1994', while a f actor of 3.3 will extend it to 2007.
d to 1999.
flux reduction based on the FP&L data is 2.63 which correspon s gq Summary and Conclusion _
l ff to evaluate An audit calculation was perfonned by BNL on behalf of the stabs l
the performente of the proposed part-length burnab e a vessel. The methodolgy respect to fast neutron flux reduction to the pressure thesis. Based on data employed by BNL was based on three dimensional flux syn t the max l'
[
supplied by Florida Power and Light it was estimated thaAssum L
duction was by a factor of 2.63.
criteria until 1999.
1.
. both units to meet the PTS scre:
Princioal Contributo_r:
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'l WCAP-11138 WESTINGH0USE CLASS 3
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CUSTOMER DESIGNATED DISTRIBUTION 1
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' REACTOR CAVITY NEUTRON MEASURLMENT PROGRAM FOR' FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNIT 3 1
n'
-1 l,
. ;L S. L. Anderson L
A. H. Foro E. P. Lippincott April 1986 l
l Work perforwed under Shop Order No. FJVP-450 and FIUP-450 l^
l APPROVED:
&A F. L. Lau, Manager l.
Radiation and Systems Analysis L
Prepared by Westinghouse for the Florida Power and Light Company
'l Although information contained in this report is nonproprietary, no distribution s' hall be made outside Westinghouse or its licensees without the customer's approval.
l I
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WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230
- 3618e:1d/050586 y
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SECTION 1 J$
PROGRAM OVERVIEW f;
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1 -1.
INTRODUCTION
- f The Reactor Cavity Neutron Measurement Program at Turkey Point Unit 3 is y
designed to provide a mechanism for the long ters monitoring of the neutron -
exposure of those portions of the reactor vessel And vessel support structure
'p which may experience radiation induced increases in reference nil ductility i-transition temperature (RTNOT) over the nuclear power plant lifetime. When
- i used in conjunction with dosimetry from internal surveillance capsules and with the results of neutron transport calculations, the reactor cavity dosimetry allows the projection of embrittlement gradients through the reactor
[
vessel wall with a minimum uncertainty. Minimizing the uncertainty in the neutron exposure projections wil'1, in turn, help to assure that the reactor L
can be operated in the least restrictive mode possible with respect to 1.
10CFR50 Appendix G pressure / temperature limit curves for normal heatup and cooldown of the reactor coolant system.
l 2.
Emergency Response Guideline (ERG) pressure / temperature limit curves.
3.
Pressurized Thermal Shock (PTS) RT screening criteria.
NOT In addition, an accurate measure of the neutron exposure of the reactor vessel I
l and support structure can provide a sound basis for requalification should operation of the plant beyond the current design and/or licensed lifetime prove to be desirable.
1 -2.
BACKGROUND 1
Over the lifetime of a nuclear power plant, changing fuel management schemes can result in significant changes in both the magnitude and distribution of 3618e:1d/050586 1 -1 1
....-l---
_~
>n;utron flux and, hence, neutron fluence throughcut the reactor vessel beltline region.
In order to accurately assess the long-term ef fects of neutron irradiation on reactor vessel materials properties, these changes in radiation level must be well known.
Each' operating reactor currently has a reactor vessel surveillance program usually consisting of from six to eight surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor-vessel wall.
The neutron dosimeters contained in these surveillance capsules provide measurement capability at a single location within the reactor y
i.
geometry.
By themselves they cannot provide the gradient information that is l
required to evaluate the impact of fuel management schemes (such as the incorporation of low leakage loading patterns) which may result in radical changes in neutron flux distributions f rom cycle to cycle.
l Additional!infonnation can be obtained by the use of supplementary passive neutron dosimeters installed in the reactor cavity annulus between the reactor l
(
vessel wall and the primary shield.
l This dosimetry package provides spectral coverage sufficient to allow the l
determination of fast neutron exposure parameters in terms of both neutron j
fluence (E > 1.0 MeV) and iron displacements per atom (dpa). The results of I
this program will establish the azimuthal and axial gradients of f ast neutron flux and dpa over the beltline region of the reactor vessel, and will provide a verification of the ability of neutron transport analyses to predict through-wall embrittlement gradients, t
l l
l 1-3.
TECHNICAL DESCRIPTION i
To achieve the goals of the Reactor Cavity Neutron Measurement Program two types of measurements are made.
Comprehensive sensor sets including i
radiometric monitors (RM) and solid state track recorders (SSTR) are employed at discrete locations within the reactor cavity to characterize the neutron energy spectrum variations axially and azimuthally over the beltline region of 3618e:1d/050586 1 -2 1
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RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 3, OYCLE'10 o-February 1985 I
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Edited by t
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1.0 INTRODUCTION
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SUMMARY
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~1.1 Introduction
.This report presents an evaluation for Turkey Point 3, Cycle 10, which
' demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was accomplished utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety.
Evaluation Methodology"II)
Turkey Point Unit 3 is operating in Cycle 9 with 56 Westinghouse optimi:ed fuel assemblies and 101 Westinghouse 15x15 low parasitic
'(LOPAR) fuel assemblies.
For Cycle 10 (expected startup June 19,1965) and subsequent cycles, ft is planned to refuel th,a Turke/ Point Unit'3 I
- core with Westinghouse 15x15 optimi:ed fuel assembly (OFA) regions.
In a licensing submittal I2) to the NRC, agproval'was requested and later approved for-the transition from LOPAR fuel to 0FA and associated proposed changes to the Turkey Point Units 3 and 4 Technical Specifications.
The licensing submittal justified the comoatibility of Optimi:ed Fuel-Assemblies (OFAs) with LOPAR fuel assemblies in a mixed-fuel core as well as a. full 0FA core. The licensing submittal contained. mechanical, nuclear, thermal-hydraulic, and accident evaluations which are also applicable to the Cycle 10 safety evaluation. Approval of the license application for the OFA transition was granted by the NRC in a SER(3) dated December 9, 1983.
l All of the accidents comprising the licensing bases (2,7) which could potentially be affected by the fuel reloao have been reviewed for the Cycle 10 design described herein.
The results of new analyses are l
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EXHIBIT
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ltl AE00/C401 LOW TEMPERATURE OVERPRESSURE-EVENTS AT TURKEY POINT UNIT 4-Case Study Report Reactor Operations Analysis Branch 9
Office for Analysis and ' Evaluation of Operational Data March 1984 i
Preparec ty:
Wayne D Lanning NOTE:
This report documents results of study completec to cate by the Office for Analysis and Evaluation of Ooerational Data with regard to a i
particular operational situation.
The findings and recommencations co not necessarily represent the position or recuirements of the i
responsible program office nor the Nuclear RegJlatory Commission.
8404050445 040321 PDR ADOCK 05000251 S
PDR 50
1.0 INTRODUCTION
Before 1979, 30 reported incidents occurred in 'pressuri:ed water reactors (PWR;)
where the pressure / temperature limits contained in the technical specifications for the reactor coolant system were exceeded.
Most of these events occurred during reactor startup or shutdown when the reactor coolant system was in a water solid condition, i.e., no steam or gas space in the pressurizer.
Over-pressure events primarily resulted from the loss of letcown flow with continued charging flow, inadvertent safety injection, or a heatup transient caused by starting a reactor coolant pump with the secondary coolant system temperature higher than the primary temperature.
These events were caused by either ecuipment malfunction or operator error.
Low temperature overpressuri:ation (LTOP) was cesignated a generic issue because f
j of the possibility.of a vessel failing by the brittle fracture mechanism.
This failure mode may be a consequence of a pressure transient after the vessel material toughness nas been reduced due to irradiation, effects (i.e., i.ncrease in nil-ductility transition temperature) while a critical si:e flaw exists in the vessel wall.. NRC rasolved the generic issue in 1979" by recommending that PWR licensees implement procedures to reduce the potential for overpressure events and install equipment modifications to mitigate such events.
Since that time, ten pressure transients have been reportec.
The two events at Turkey Point Unit 4 on November 28 and 29, 1981 exceeded the technical specification limit (415 psig below 355'F) by about 700 and 325 psi, respec-tively.
The two events were. designated Abnormal Occurrences by the NRC (Ref. 1).
The other eign: reported events were mitigated by the overpressure protection system.
These two overpressure events anc a significant numeer of events at i
other PWRs involving inoperaele trains Of the overpressure protection system promoted AE00 to initiate an evaluation of operational events with the focus primarily on Turkey Point.
.L The overpressure protection system and the overpressure events at Turkey Point Unit 4 are cescriced in Sections 2 and 3.
Section a contains the analyses anc L
evaluation ~of the two events, including utility management's reaction to the events.
Section 5 reviews the operational experience related to inoperable trains of the overpressure protection system at otner PWRs.
Section 6 evaluates the aceocacy of existing LTOP technical specifications.
Section 7 ciscusses j
the need fer coerating in a -ater solic concition.
Sect'or S lists the finc-l ings anc conclusions, anc Section 9 contains the I.E00 -ec:mmencations cased on tnts case stucy.
(
I "NUREG-022a entitlec,"ReactdrVesselDressureTransient i
Protection for Pres-t sur;:ec Water Reac*. ors,"
as cu0lisnec in Septemcer 1975 cocumenting the com-O pletion of the generic activity.
LTCP mitigating systems were installed in 1
most plants beginning in 1979.
1 f
l 1
hl
h
~
^ ^^
1.0 INTRODUCTION
AND $UMMARY
1.1 INTRODUCTION
This report presents an evaluation for Turkey Point Unit 4, Cycle 10, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was accomplished utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology"(I).
c Turkey Point Unit 4 is operating in Cycle 9 with all Westinghouse 15x15 low parasitic (LOPAR) fuel assemblies. For Cycle 10 (expected startup mid 1984) and subsequent cycles, it is planned to refuel the Turkey
-l Point Unit 4 core with Westinghouse 15x15 cptimized fuel assembly (OFA)
In a licensing submittal (2) to the NRC, approval was regions.
requested for the transition from LOPAR fuel to 0FA and associated proposed changes to the Turkey Point Units 3 and 4 Technical The licensing submittal justifies the compatibility of Specifications.'
OFAs with LOPAR fuel assemblies in a mixed-fuel core as well as a full The licensing submittal contains mechanical, nuclear, 0FA core.
thermal-hydraulic, and accident evaluations which are applicable to the Approval of the license applicat1cn(2) t Cycle 10 safety evaluation.
I3) dated for the OFA transition was granted by the NRC in a SER December 9, 1983.
In a separate licensing submittal
} to the NRC, approval was I
limit to 1.62 at normal requested to increase the maxirr.um F3g operating conditions as part of a vessel flux reduction program (5) g, The report partially resolve the pressurized thermal shock concerns.
contains nuclear, thermal-hydraulic, and accident evaluations which are Approval of the license applicable to the Cycle 10 safety evaluation.
application ( ) for the increase in the F 11mit was granted H
by the NRC in a SER(6) dated December 23, 1983.
~
1 l
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.i, 10 INTRODUCTION AND
SUMMARY
j i
1.1 Introduction This report presents an evaluation for Turkey Point 3 Cycle 11, which
' demonstrates that the core reload will not adversely affect the safety of the plant., This evaluation was accomplished utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology"II)
Turkey Poin't' Unit 3 is operating in Cycle 10 with 112 Westinghouse optimized fuel assemblies and 45 Westinghouse 15x15 low parasitic (LOPAR) fuel assemblies. For Cycle 11 (expected startup mid-May, 1987) and subsequent l-cycles,itisplannedtorefueltheTurkeyPointUnit3corewithWestinghouse I) 15x15 optimized fuel assembly (OFA) regions.
In a. licensing submittal to the NRC, approval was requested and later received for the transition from LOPAR fuel to 0FA and the associated proposed changes to the Turkey Point Units 3 and 4 Technical Specifications. The licensing submittal. justified the L
compatibility of Optimized Fuel Assemblies (OFAs) with LOPAR fuel assemblies L
in a mixed-fuel core as well as a full 0FA core.
The licensing submittal contained mechanical, nuclear, thermal-hydraulic, and accident evaluations which are also applicable to the Cycle 11 safety evaluation.
Approval of the l
license application for the OFA transition was granted by the NRC in a SER(3) dated December 9, 1983.
l A significant number of Integral Fuel Burnable Absorber (IFBA) rods are being l.
L used for the first time in Turkey Point Unit 3* as part of the Region 13C and l'
130 fuel assemblies. These rods are described in Section 2.1.
A more detailed description and evaluation of IFBAs for 14x14, 15x15 and 17x17 fuel arrays are given in References 4 and 5.
The NRC has approved the use of IFBAs 4
for Westinghouse fuel rods in 15x15 fuel assemblies (6) u L
- Turkey Point Unit 3 did have demonstration IFBA rods in Cycles 8 and 9.
,wwsro n 1
p?
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~~Co:
j
'3"'
R.sr-0054 8
t.-
y RELOAD SAFETY EVALUATION TURKEY POINT PLANT UNIT 4, CYCLE 11 REVISION.1 April 1955 t
Edited by:
J. S. Baker J. Skaritka
)
1' i
1,.
Approved:
g,W
[.#',gAw E. A. Drenis, Manager 9 i-Core Operations 5
Nuclear Fuel Division
[
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.... I.
- _... _ _..... - ~,
^-
b
1.0 INTRODUCTION
AND
SUMMARY
l g,1. INTRODUCTION i
This report presents an evaluation for Turkey Point Unit 4 Cycle 11 demonstrates that the core reload will not adversely affect the safe This evaluation was accomplished utilizing the methodology described 51) plant.
in WCAP-9273, " Westinghouse Reload Safety Evaluation Nethodology" Turkey Point Unit 4 operated during Cycle 10 with 117 Westinghouse 1 d fuel parasitic (LOPAR) fuel assemblies and 40 Westinghouse 15x15 optimiz l
For Cycle 11 (expected startup Nay 1986) and subsequen';
assemblies (OFA).
cycles, it is planned to refuel the Turkey Point Uni.t,4 core with prim In a licensing Westinghouse 15x15 optimized fuel assembly (OFA) regions.
submittal (2) to the NRC, approval was requested for tho' transition from LOPAR fuel to 0FA and associated proposed changes to the Turkey P h
The licensing submittal justifies the and.4 Technical Specifications.
compatibility of 0FAs and LOPAR fuel assemblies in a mixed-fuel core I-The licensing submittal contains mechanical, nuclear, as a full 0FA core.
thermal-hydraulic, and accident evaluations which are applicable to the Approval of the license application (2) for the OFA 11 safety evaluation.
transition was granted by the NRC in a SER(3) dated December 9, 1983.
1 In a separate licensing submittal (4) to the NRC, approval was requested to limit to 1.62 at normal operating conditions as increase the maximum Fg part of a vessel flux reduction program (U) to partially resolve the l
The report contains nuclear, pressurized thermal shock concerns.
thermal-hydraulic, and accident evaluations which are applicable to the C Approval of the license application (4) for the 11 safety evaluation.limit was granted by the NRC in a SER(6) dated increase in the F H j
December 23, 1983.
1 adadt 6 400d27 a,
a a
L
' Steel Hector & Dayb.
a Me,m. no,t
[ Mt m T, W '
4-c30s) sn.se3e L
October 13, 1989 f.
Joette Lorion-y center for Nuclear Responsibility
[J '
'S901 S.W.
74th Street Suite 16304 South Miami, Florida 33143 7
. ~.
Re: Florida Power & Light Company (Turkey Point Plant, r
Units 3 and 4), Docket Nos. 50-250-OLA-4 and l
50-251 OLA-4 (P/T Limits)
. i.
l,
Dear Joette:
l I am enclosing copies of the safety evaluations for the Unit 4, Cycles 10 and 11 fuel reloads..Together with the safety evaluations previously delivered to you, you should now have the safety. evaluations for Unit 3, Cycles 9, 10 and 11, and for Unit 4, Cycles'.10, 11'and 12.
These represent the evaluations ~for cycles that. covered'the period beginning in'1985 and. extending to the_present.
t 4
You'also asked me for'the capacity. factors for years prior'to 1985.
I believe the following is responsive to your I
request (1974 was the first year for which the information was i
I available to me):
g 1'
Unit 3 Unit 4 L'
I j
1974 62.1 74.1 1975 75.0 68.4 1976 73.8 64.5 l
1977 76.6 62.8 l-1978~
77.1 64.9 l
l 1979 49.3 65.9 i
l-1980 77.3 67.9 1
1981 16.1 78.5 1982 66.5 67.9 t
+
1983 75.0 51.7 l
l 1984 81.8 52.6 i
My records reflect that you now have all the information you requested.
Please contact me if this is not P
1 ggg I
L'4 41#r Merm Omco 1200 Norintmd0e Centre 1 440 Rovel Pam Way 1200 Corporate Piece 201 Souin Monroe 4000 Soumeest Francies Center West Pern Beach FL 33401 4307 Pern Bearn. FL 33480 1200 Norm Feoeres Nignway Tenaneaeus. FL32301 1648 Merm. #L 33131 2396 (335) 860 7200 (305) 650 7200 Boca Reson. FL 33432 (904) 222 4194 (306) 677 2800 Far (306) 656 1500 (305) 394 5000 Fax (904) 222 8410 Fer (305) 358 1418 Far (306) 394 4856 d
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'Joette Lorion.
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i October'13, 1989 Page 2-E your_ understanding as.well' I apologize for the earlier confusion and hope that, by providing the missing information to you within a day of your request, I have avoided any serious incenvenience on.your part.
Sincerely,
[V j.
John T. Butler sEnclosures-cc: Steven P. Frantz, co-counsel for Florida Power & Light Company i
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c Of fice of 'auclear teact:r Regulation l
'v. Carrell 3.11sennut. Director
. Attention:
Civt<Jicn of Licensing
'J. S. Nuclear Regulat:ry Ox. mission
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4tnington. 2. C.
20!!$
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Turkey N int "afts 3 and 4 l
Oc ke* Mos. 50 250 in-50-251 proposes Li:anse u.erc.e.ent-siant !.eveti'ance da:ec al
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- n ac:or:a~. e.ita.0 ~FR 50.70. Florida P wer 1 Lt;nt Cx:any s
.<..reui ta ter te sig e:
a?;eecix A :f Tsetlity 0;ersting icenses ;P8 31 and 41.
Ihis amencvt is ;re;ose to ::rneine t".e estct:r materials surveillance l
.ait 3 and a f ato a single integetteo program.ni:n c:nfor ss i
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- rogram 3
to One recu' 1rnents of 10 *FR 50 s :encices 3 and H.
l anc snown on :ne tc::mpanying The 3r:00ste n endment is escrioed below l
7ecani: 31 !:ecificaticn : ages.
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Id ol e 4. 2 *. 2 *o I nge 4. 20 1 The Irradia**:n S;ecimen Ichedule (itern 7 2) in Iaole 4.21. is deleted and a etvised ven':n to reflect the Dr:0osed intejrlted progr3m is added 12 Page 1.20-1.
delete 3 4.2 11 S t e e s B 3. '. - 3. ? 1. !.*. 2, 3 A. 2 13. B 4. 20-1 t r*:
4 tn the Jeove changes art revised.
- Me :asas associttM l
the Turxey Point Plant Nucteer Ne :recosee mencmen: 945 been reviewed of C:n#.:n and t. e Florida Power 3 Ligne Company luclear teview Scard.!
Safat/
F:' recuesu issuance of this ;rcoosed amenc.med i
or:er to Al':= ;rocer imolementation of tne single integested program,
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In ac:ordanen d:n 10 CFt 50.91(b)(1), a :ocy of the prooosed amendm for tne State of Florida.
- eing for.ae.es to the State Cesignet 1p-u
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pe vita 10 CFR 170.12(c), a chec.t for $150 is attached,
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d action in light of tne three ggp is an evaluation of the pro:eseCFR 50.92 (H,11ptficant Hazar:Is),
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V-notch :oecimens, ten Charpy specime.s g:n Ty;e 1 esosule contains 23 Charoy The remaining eight CSL-;y
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.nachined t-om each of the two shell f:rgi gs.
h T. se 1 specimens 3.re machined from correlsted moni:x material. In addit eet!
7 esosule centains four tensile specimens (two s:ecimens from eacn 2ng f:rgints) 193 six TCL seee! mens (three specimens fro i dd iddle. And aluminum-c: salt wire are secured in holes erilled in spacers at tne too f: gings).
l bottom of sich Type i es:sule.
i eight specimens Tyx *1 cs:sule c:n: sins 32 Charpy V-noten specimens:
l d eight mac.ined it:m :ne of :ne s..e!! f:rgings, eigtt cecimens of weld meta an l i men!::ts.
Esch s:ee!me :s :f HC mets) : e remsining eignt cecimens are corre at on d four T0t.
- n ac:1:t., eacn Tv:e !! es:sule c:ntain.s !:cr tensi!e ce f the shel!
s f:rging s..d :ne. eld - e:11. Esen Tvpe !! :cs:mium-ou:Meatice esesuies c:ntaining s ec:me.s 2:
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- enter of :.e :s:sL!c.'.se::;es ::atium- ?3 andmeetuntum-237, are cents
- ---:s!nment s!!:r:f ed by :ne mi-eter assembly trevents loss and
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- ntsmi.-2:icn ty :ne.e:t.nium-237 and.r:.n.ium-;?3 an.d ineir sc::va:
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mi;;igts- :s :f ursnium. }$ ::ntainec in a 37!.inen-(
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s.-d is misced M 2 !! inen cismeter hole in tr.e :esimeter blec'< one.ee uarou
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located 1: :ne snietee3 La-;num c:esit tre s!so secured.n -oles tri!!ed ;n spacers
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8 TURKEY POINT UNif5 3 AND 4 REACTOR SURVEILLANCE eAf!R!AL P40 GRAM PROPOSED CHANGE TO PLANT 7!!dN! CAL $PE*! FICA!!ONS o
1;;endia H re;uires reactors constructed of ferritic materielt nave their j
teltline rt;*cas monitored by a surveillahre program conplying with ASTM
'[18$.
Appensis 3 defines beltline materials as shell material inclucin; = elds and neat af fected :enes, plates or forgings, that directly surround tne effective "**;nt of the fuel element assrelles.
ine existia; Turkey Point 3 and 4 surveillance programs contain t.o ty;es of surveillance :a:svles:
5 T :e I capsules t:ntain forging simples only; 3 f pe
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- ' :sosul ts ::etain forging, wels, anc ha:
f
- samples, i
i Se first T :e II *a:Sule re90ved has cefi9e3 the most l'ai*ing material in I
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{ t e tactor as tne gifta -e'3s basec on feteture toughneta requirements.
I attacnment 1 i t an ex:ea;;
- 3?. tne 8TP s.meillance program.
i re.s t e nu= car an: icent4catier. syttes ;f Type I and !! :apsules in esca i
- ( t 9 e T ar t r/ 8: int Vessels, t s c a n : e t w**. * * * *, a r e on tf t.c Tj;e ;I i
- 5 4I t s r "mJ ' ' ' - *
> *-* el. ~Attac ament 3 shows the c aosv e e.gca ti ons,
- ::tJ'1 t e m:st einin;*;' *esults frst 19e existing ;c ;r Am and to.:.:ste
- e :P:gr a' :: tar e9: ' :ta:'t H requi*emeats. FPL pro oses to remove da'
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- :t * '. s a* t'111"P.e ca:s;l es f 3r the remai :er of plant li fe.
This et:si!es
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t*st 3 ta:s;* es :e tvaila:le far removal ter? ugh the end of life.
Since taere t e :mi f 2 :::s;les avail 31e for each unit, e propose to integrate tne 1.*re ' ' i m a: e : :;s tms a s e -mi t t ed >y Ap pe *: 4 4 H, !!, C.
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??> 3 and a are identical in :esign, share icenti:al 81 ant Technical Speci fications anc inte nac identical major cg todifications sucn as steam generator repincement anc TMI sect fit modi fic a ti on s.
The reactor vessels were fa:ricatec :ne same ay i
ty the same su:: lier utili:19; :ne same material s.
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Nterials 612. SA 302 grace B ferritic steels {ll reactor m swil girta -elds were mace t the intermeatate g3*i
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ire;(heat No. 71249) and ice :s fied as SA1101
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3-is the material =ith the hi;nes ;rectetes Ri for both units Since tnis gis particularly.e,ll suited to an integrated surveillit is o)
- )
- edicted Severity of Irradiation ance progrwe, 33th reactor vessels are expected to es;erien e an end
%en:e :f a maxi,um of 1,81;H of life j
- er3:ec asin3 similar fuel i:a::n;:3f 32 (p 1.ev) and %ve t
sinca sta-up, B e *ur!
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.all vessel n en:a pent:;iens 37,.
'or 'init 4 fn A:-* 1 1-15 ';r Units 3 4 :
The : 'fe-. ce is :ve to ; f recen; 3,79 g, ppy1%3 in Oct
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' e '. ave installed 9t: Ort dosime:'y tr3und both IJ-itv Poi vessels t3 Otachmark in0ivicga5 sy:le fl. gen:g n;
- .r :e:eacence On incare survei'1ance :apssie f ii :ssi, ;3,re3y 2:
att Plais; 3e t.een pl ants met f.
3 0 *. *
.Miss havt ::mm:n managesent, tag :ne surveillance pro
'a**;t
- / :*e Dces a90 In30ec;4
.e:t-en:
7s se: tion of the qu lege [3g.;y tre sta r f, I:
{:*:' *;t':/ :lan in the !<ent of te:'J:td Po-er Operati:ns
..:a;e 3Ct' 0' ants have caosules, 4
5*:st.antial Advantages To Be Gaine:
o Se.ain ad<antage is 00:aining :ne :es:
~3054It removal.
da:a availaole from en:3 Accitional acvanu;e will be realized-from fewe
}00-s:Fle renovels and toth plants :'
2 0 ="1 Ert s sur? t eatpeat *Urt 0*;*vt s,4 at197 to identical neat us and '
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i jLa2' April 11, 1977 i
J' L-77-113 I(,
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Office of Nuclear Reactor Regulation Attention:
Mr. George Lesr, Chief
.h-{ 'Qli, ( Jif C.*
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.1 Operating Reactors Branch 43
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({" s,i,T g, l
Washington, D. C.
20555 W,.
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Turkey ;oint.nis 4 3.
C :ket No. 50-U.1
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*-+u.re Touchne22 E ruira ents o
l1 On April 7, 1977, a n ::ing was held *ith nar.bers of your ataff to dis:uas the sta us r f the Turkey 7: int Unit 4 rea: tor
$ j vessel with respect to :he fracture :::;hness requirements y l of 3ection V.B of Appanfix G to 10 TE 50.
At that meeting, 2
A we showed that the veld metal surveillance data-for<thc.Tarkey
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7 int Unit 3 reacter ":ssel represen: not oniv the.co-L i
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i Csta supporting this conclusi:n are a :2ched.
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The data show that the valdment sanp'.es f rom a Unit 3 sur-a
?g veillance capsulo "T" and fr:n both the Unit 3 and Uni: 4 reactor vessels were mafo frem the ss..e combination of filler
,y
- P wire heat number and welding flux ist number.
Scuever, tha 1
weld:.ent samples f rem a Unit 4 survei'. lance capsulo "T",
although g
I containing the sam 2 filler wire hea " number, 'used V different' l
.s L$
welding flu >: lot numhcr..Therofere, the Unit 3 capsule "T'
samplo is more represen:stive of the. nit 4 reacter vessel.
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,lg Irradiation data f rom the Unit 3 capsule was submitted to tho c::hibited a shole MRC on Octcber 19, 1976 (L-75-363).
Thedatg3 l q onergy of 53 f t-lbs at a fluence of 5.7 :: 10 nv' Accordingly, l
7.,
the mid-plane circunferential vascal, eld in Unit 4 can bc l
c::pected to naintain a shelf energy
'.c rel in e:.:cesa of 50 f t-lbs I
at the 1/4 T location until at less: June 10$0at.thicg3 time '
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f this location will have received a fl.:ence of 5.7 x 10 nyt.
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t In the October 19 letter, we also stated that additional r.2,.
were being prepared by our :tsSS vender to :omplete summart:
and fracture analyses for Units 3 an..
p.-
the fatigue, accident, We expect to receive thase additional reperts in draft form i
about one week, and should be able to forward the.m on to yc..
office in approximately 6 to 8 weeks.
The evaluation discussed above supports the conclusion we,,
i t
presented at the April 7 meeting that 'an Appendix G inserv L..
i inspection of the_ Unit 4 reactor vessel helt-line area nee.t j
not be conducted until af ter June.1990, i
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very truly yours,
- ---- ]
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chert E.
Uh N e vice President EC/MA3/cyc Attachment,
.Mr. Norman C. Moseley, Region II cet Robert Lowenstein, Isquire
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i LOW TEMPERATURE OVERPRES$URE i
EVENTS AT TURKEY POINT UNIT 4 t
6 l
Case Stucy Report Reactor Operations Analysis Branch t-i Office for Analysis and ' Evaluation of Operational Data March 1964 s
Preparec ty:
keyne D. Lanc.ing NOTE:
This report cocuments results of stucy completec ta cate by the Office for Analysis ano Evaluation nf Ooerational Data with regara to a particular operational situaticn.
The findings and recommencations do not necessarily represent the p?tition or reovirements of the responsible program of fice nor the Nuclear Regulatory Commission.
t 8404050445 840321 PDR ADOCK 05000251 gj S
=
- 1. 0 INTRODUCTION gefore 1979, 30 reported incidents occurred in pressuri:ec water reactors (PWRs)
I where the pressure / temperature limits Contained in the technical specifications for the reactor coolant system were exceeded.
Most of these events occurred during reattor startup or shutdown when the reactor cociant system was in a water solid condition, i.e., no steam or gas space in the pressuri:er.
Over-pressure events primarily resulted from the loss of letcown flow with continued charging flow, inadvertent safety injection, or a heatup transient caused by p
starting a reactor coolant pump with the secondary coolant system temperature higher than the primary temperature.
These events were causec by either equipment malfunction or operator error.
Lew temperature overpressuri:ation (LTOP) was cesignatec a generic issue because of the possibility of a vessel failing by the brittle fracture mechanism.
This failure moce may be a consecuence of a pressure transient after the vessel material toughness has been recuced cue to irradiation, effects (i.e., increase in nil-ductility transition temperature) while a critical si:s flaw exists in the vessel wall.
NRC resolved the generic issue in 1979* by ricemmending that PWR licensees implement prosecures t.o reduce the potential for overpressure events anc install equiptrent modifications to mitigate such events.
Since that time, ten pressure transients have been reportec.
The two events at Turkey Point Unit 4 on Novemeer 28 and 29,1921 esceecec the technical specification limit (415 psig below 355'F) by about 700 anc 325 psi, respec-t,i v e l y.
The t o eventr..ere :esignated Abnormal Occurrences by the NRC (Ref. 1).
The other eignt repor*ec events were mitigated by the overpressure protection
- system, inese t-o overpressure events anc a significant numcer of events at other PWRs involving inopera:1e trains :f the over; essuee :r:tection system prometec AE;t to initiate an evaluation of operational everts with the focus
- rimariiy :n T.ney Point, The overpressure protection system and the overpressure events at Turkey Point Unit a are cescrieec in Sect'ons 2 and 3.
Section a contains the analyses anc evaluation of the two events, including utility management's reaction to the events.
Section 5 revie-s the operational experience related to inoceraDie trains of the overpressure protection system at other N Rs.
Section 6 evaluates I
the aceculty Of esisting LT07 technical s p e c i f i c a *. i o n s.
Section 7 ciscusses the neec f:r ::erating in a -ater solic corcitten.
Sect *:n i lists tne fin:-
i ings anc conciusions, anc Section 9 c:ntains the ;E;; e::vencations :asec on l.
this :ase stucy.
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- NU R E G - 00 2 t. entitlec, "teact:- Vessel Dressure Transient Pr:tection for Fres-suri:ec water Reacters," -es cuelisnec in Se: tem:er 1975 cccumentsn; the ecm-e 1etion of the generic activity.
LTCP miti;ating syste+s -e e insta11ec in l
most plants beginning in 1979.
1 m __ _.~
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1.0 INTRODUCTION
AND $UMMARY l
1.1 INTRODUCTION
This report presents an evaluation for Turkey Point Unit 4. Cycle 10, j
which' demonstrates that the core reload will not adversely affect the l
safety of the plant. This evaluation was accomplished utilizing the l
I methodology described in WCAP 9273, " Westinghouse Ratoad Safety Evaluation Meshodology"(I) t j
Turkey Point Unit 4 is operating in Cycle 9 with all Westinghouse 15x15 low parasitic (LOPAR) fuel assemblies. For Cycle 10 (expected startup l
mid 1984) and subsequent cycles, it is planned to refuel the Turkey i
Point Unit 4 core with Vestinghouse 15:15 optimited fuel assembly (OFA)
In a licensing submittal (2) to the NRC, approval was l
regions.
requested for the transition from LOPAR fuel to 0FA and associated proposed changes to the Turkey Point Units 3 and 4 Technical The licensing submitta) justifies the compatibility of j
$pecifications.
CFAs with LOPAR fuel assemblies in a mixed-fuel core as well as a full l
The licensing submittal contains mechanical, nuclear, CFA core.
l thermal-hycraulic, and accident evaluations vehich are applicable to the Approval of the license application (2)
Cycle 10 safety evaluation.
I3) 4ated for the OFA transition was granted by the NRC in a $ER l
December 9, 19P3.
In a separate licensing submitta1I#) to the NRC, approval was f'
limit to 1.62 at normal requested to increase the maximum FAH i
operating concitions as part of a vessel flux reduction program (5) g, The report partially resolve the pressurized thermal shock concerns.
contains nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 10 safety evaluation. Approval of the license f
N limit was granted application (
for the increase in the F g by the NRC in a $ER(6) cated December 23, 1983.
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1.0 INTRODUCTION
AND
SUMMARY
i 1.1 Introduction l
i This report presents an evaluation for Turkey Point 3, Cycle 11, which l
demonstrates that the core reload will not adversely affect the safety of the l
plant.. This evaluation was accomplished utilizing the methodology described in WCAP-9273, ' Westinghouse Reload Safety Evaluation 4 thodology*II) l Turkey Poin't Unit 3 is operating in Cycle 10 with 112 Westinghouse optimized l
l fuel assemblies and 45 Westinghouse 15x15 low parasitic (LOPAR) fuel r
assemblies.
ForCycle11(expectedstartupmid-May,1987)andsubsequent i
cycles,itisplannedtorefueltheTurkeyPointUnit3corewithWestinghouse I) 15x15 optimized fuel assembly (OFA) regions.
In a licensing submittal to the NRC, approval was requested and later received for the transition from LOPAR fuel to 0FA and the associated proposed changes to the Turkey Point i
Units 3 and 4 Technical Specifications. The licensing submittal justified the compatibility of Optimized Fuel Assemblies (OFAs) with LOPAR fuel assemblies f
in a mixed-fuel core as well as a full 0FA core. The licensing submittal contained mechanical, nuclear, thermal-hydraulic, and accident evaluations which are also applicable to the Cycle 11 safety evaluation.
Approval of the l
license application for the OFA transition was granted by the NRC in a l
SERI3) dated December 9, 1983.
l A significant number of Integral fuel Burnable Absorber (IFBA) rods are being used for the first time in Turkey Point Unit 3* as part of the Region 13C and 1
h 13D fuel. assemblies. These rods are described in Section 2.1.
A more detailed deso iption and evaluation of IFBAs for 14x14, 15x15 and 17x17 fuel arrays are given in References 4 and 5.
The NRC has approved the use of IFBAs for Westinghouse fuel rods in 15x15 fuel assemblies (6) i
- Turkey Point Unit 3 did have demonstration IFBA rods in Cycles 8 and 9.
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RELOAD SAFETY EVALUATION TURKEY POINT PLANT i
UNIT 4, CYCLE 11 REVISION 1 i
1 April 1955 9
li I
t Edited by:
J. S. Baker
[
J. Skaritka t
Ipproved:
472 o ',d A n E. A. Orenis, Kanage#
Co e Operations Nuclear Fuel Division
,,p gXHittflT' 9
. :.... w 2:
I
1.0 INTRODUCTION
AND
SUMMARY
g,1 INTRODUCTION This report presents an evaluation for Turkey Point Unit 4 Cycle 1 demonstrates that the core reload will not adversely affect the saf This evaluation was accomplished utilizing the methodology describe (I) plant.
in WCAP-9273, ' Westinghouse Reload Safety Evaluation Methodology" Turkey Point Unit 4 operated during Cycle 10 with 117 Westinghous fuel parasitic-(LDPAR) fuel assemblies and 40 Westinghouse 15x15 op l
For Cycle 11 (expected startup May 1986) and subsequent assemblies (OFA).
cycles, it is planned to refuel the Turkey Point Unit,4 core with pr In a licensing Westinghouse 15x15 optimized fuel assembly (OFA) regions.
submittal (2) to the NRC, approval was requested for tho' transition from LOPAR fuel to 0FA and associated proposed changes to the ' Turk The licensing submittal justifies the and 4 Technical Specifications.
ll compatibility of 0FAs and LOPAR fuel assemblies in a mixed fuel core as The licensing submittal contains mechanical, nuclear,l as a full 0FA core.
thermal-hydraulic, and accident evaluations which are applicable to th Approval of the license application (2) for the OFA 11 safety evaluation.
I3) dated December 9, 1983.
transition was granted by the NRC in a SER
- 54) to the NRC, approval was reqcested to in a separate licensing submitta1 limit to 1.62 at normal operating conditions as increase the maximum F3g part of a vessel flux reduction program (5) to partially resolve the The report contains nuciaar, pressurized thermal shock concerns.
l thermal-hydraulic, and accident evaluations which are applicable to Approval of the license application (4) for the 11 safety evaluation. limit was granted by the NRC in a SER(6) dated N
increase in the F3g j
December 23. 1983.
1
.o n......
steelHector& Davis e mina.
m t. D asr posi m. sos.
3 October 13, 1989 Joettc Lorion conter for Nuclear Responsibility 5901 s.W. 74th Street suite #304 South Miami, Florida 33143 Ret Fluida Power & Light Company (Turkey Point Plant, Unite 3 and 4), Docket Nos. 50-250-OLA-4 and 50-251 OLA-4 (P/T Limits)
Dear Joette:
I am enclosing copies of the safety evaluations for the Unit 4, Cycles 10 and 11 fuel reloads.
Together with the safety evaluations previously delivered to you, you should now have the safety evaluations for Unit 3, Cycles 9, 10 and 11, and for Unit 4, Cycles 10, 11 and 12.
These represent the evaluations for cycles that covered the period beginning in 1985 and extending to the present.
You als0 asked me for the capacity factors for years prior to 1985.
I believe the following is responsive to your L
request (1974 was the first year for which the information was available to me):
Unit 3 Unitj 1974 62.1 74.1 1975 75.0 68.4 1976 73.8 64.5 1977 76.6 62.8 1978 77.1 64.9 1979 49.3 65.9 1980 77.3 67.9 1981 16.1 78.5 1982 66.5 67.9 1983 75.0 51.7 l
1984 81.G 52.6 My records reflect that you now have all the information you requested.
Please contact me if this is not I
b{f5.exwn m.
... - n o.o,.
,,- - n 1 - a = =,., -
1We 904 410
.. to....,.
pag)
1
/
Steel Hector & Davis
(/.
Joette Lorion i
October 13, 1989 Page 2 your understanding as well.
I apologize for the earlier confusion and hope that, by providing the missing information to you within a day of your request, I have avoided any serious l
inconvenience on your part.
]
Sincerely,
[V John T. Butler Enclosures l
cc: Steven P. Frantz, co-counsel for Florida Power & Light Company l
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Of fice of 'tuelear tetet:r Rejulttion j
- v. Carrell 3. Eisennut. 01rtetor givision of Licensing I
. Attention:
'J. '!. Nuclear Regula::ry 2xmission uasnington, C. C.
20!!5
- ear *.r. E!sennut:
Re: Turxey N'P.: 'Jnits 2 t-c 4 Oc:te: %s. !C-C!; sn: 50-251 P-coose Li:ense A.erement -
tes::sr M ar: ! rve! 'ance da: red ii : ~:r m 74 30.70. Florics 7:we 5 '.1;n Cx:any sa mits
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.:;eacia a :# Facility 0; easting.icenses :78 31 and 41.
is ;recesM to :xcise : e Pttc::e materials surveillance
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eenement is :es:ri:ed N1:= tac snown on :.e sc::m;anying The r:;os e:
Iac'.n i: 31 !:+ci ficacicn :ajes.
Isol e 1. 2.'. t ac 8 tt e 4.20 1 Icnedul e (1:en 7 2) :n Idole 4.21. is deleted ams 5
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and delete 3 1 2 11 310 e s 3_ }. '. 3. ? 1. 2.*. 2. 3 A. 2 13, 3 1. 20 1 M vi:n the Jeove changes are revitec.
~5e :ases ss:ocie:
Plant Nuclear 9as been reviewed :y t.*re fursey Poin:
Saf ety CW.ctee and One Florica Power,1 Lignt Com:any Nuclear Re St :recosac n enenen:
Mis proposed Amend.'annt :sfort *Me beginning of the 7:'. recuests issuance :f :3 refueling outage (currently scheduled to begin 3-30-4
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5: ring 1935 Jii:or:er :s alt:= :recer im:lementation of ne single integrated proge F[
e nt is
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av TURKEY POINT yi ?$ 3 AND 4 RE ACTOR SURVEILLANCI a.ATERIAL PROGRAM it0 POSED CHANGE TO PLANT ?!:dNICAL spt:! FICA!!0NS t
1:,encia M rt: wires reactors constructec :( ferritic materitts nave tr,eir teltline et;*ons monitored by a surveillance program complying witn ASTM
(;t5.
a pensis 3 defines beltline materials as snell material inclucim; neles c
and helt af *ecttd :enes, plates or forgin;s, that directly surround the e f f ec t i ve ne d ;nt o f the f ue l el et.en t a s s emel l e s.
?nr. existin; Turkey Point 3 and 4 survetilance programs contain t.o ty;es of survetilance us:sults:
5 Ty:e 1 caosules ::ntain for;ing simples onif; 3 Type
- :::sults :entain for;ing. welc, anc ha: sanples.
l
- e 'irst Tj;e !! :s:su'ie r emoved has cefi9e2 the *ost limiti.*g material in
- e enctor es the girtn se'ts base: on fr6cture toughness requirements.
I ttia:* en*
1 is an ette";; **tm the 377 s. veiItance program.
At'aC9 ment 2 1
.s e sumcar ac: 1:ent:ation systes :/ Ty:e 1 anc 11 :a:sules in ea:n
**e Tartry **ia.*, Vessels.
A s " a n : e
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esults f om :te existin; ;r:; ram and to.o:ste e : :;rr, :: :.r en: 1::ta:'t 9 resvi rse is, Fpt pr:coses to remove d 'y
- :e ::
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ts 'er the risad :er of plan:
I life.
ints re:.tres a: : :a:s ;' es :e avai'. a:'.e *:r removal treougn the enc cf i t 'e.
Since :nere t e : 'f ? :a:ssies avtila:'e for each unit we propose to integrate tne
- . <e ' t aa: e : :; t s i s :e--i t ted by A;;e-:'
d, !!. C.
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e; ee :f *:m.mena t i:y 4}
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- > 3 and 4 are icentical in :esi;n, share iden:1:ai Plant
- enni:a1 Scecifications anc are nac icentical major e
,oct rications seen as steam ;enerat:r replace ent anc Txt occtfit 90?t#iCa*i0ns.
The re!C*:e vessels were fa rica*ec One sine may by tre samt sa :'ier u t il i :19; ne same materials.
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Attention:
Mr. George Lear, Chis!
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Operating Reactors Bran:h 43 y
Divisi:n of C::eratin-Rea:: ors
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7l U. S. Muclear Regula: cry CO:/31ssicn
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- ear Mr. Leart
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Cn 'spril 7, 1977, a n:n'_ng was held n:h ner.bers of ycur ataf f dis:uss the sta:us -! the Turhey 7::.nt Unit 4 res:t2:
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- iessel with respec: to :he fra::ure :::;hness requirer.onts of 3ection V.3 of 'sp.:s.fix G so 10 *T7. 50.
At tnat *..20 tin 7, i
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A l we showed that the weld metal survei.11ance data for the Tar.' ey n: t o.n. iv.i t h e c, o.r e,...._l l '.
r 7:in: Unit 3 react:: ":ssel rerresen:
I, nidplane circumfaren-ial welds in Uni 3,- but n Un_i, as we Cata sue.rortinV this ::::1usi:n are attached.
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The data show tha: the weldr.ent san..as frem a Unit 3 sur-i d
i veillance capsule "T" 2..d fr:n both the Unit 3 and Unit 4 M)l reactor vessels were maio fr m the sar.e cc.tbinatica of filler k
vire heat n=ter and we' ding flux 10: nr3e r.
.9evever, the weidnent samples f r:m a Unit 4 survei'.*.ance capsule "T",
alth: ugh s
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containing the same filler wire hea sv..ber, used E different e
velding flux lot n=.har..Therefere. :he Unit 3 capsul.a T
M sample is more reprisen:stive of the *.* nit 4 react:r vescel, w
S Irradiation data from the Unit 3.:apsule was suhr.itted to tho MRC en Octcher 19, 1975 (L-75-363).
he data exhibittd a shelf onergy of 53 ft-lbs at a fluence of 5.7 :: 1013 nv-Ac:ordingly, the mid-plane cire'.nf ersntial vassal, eld in Unit 4 can bc l
e::pected to maintain a shelf energy *.evel in excess of 50 f t-lbs a
at the 1/4 T location until a: leas:.une13SOatwhicg3 time '
j this location will have received a filence of 5.7 x 10 nyt.
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- n the October 19 letter, we also stated that additiona), r a,.
were being prepared by cur 215SS vendor to :omplete swcmaria the fatigue, accident, and fracture analyses for Units 3 an..
7.
We expect to receive these additional re;;rts in draf t for a about one week, and should be able to fen'ard the:n on to ye..
office in approximately t> to 8 weeks.
The evaluation discussed above supports the conclusion we presented at the April 7 meeting that aW Appendix G inservl....
inspection of the Unit 4 reactor vessel belt-lina area nee.t not be conducted until after June 1990.
- /ery truly yours, Tm Q
?. chert E. Chri.D.
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..e Vice President 7.I'.'/:G3 / Op c Attac h..en t Mr. Ilorman C. Mcseley, Region II cct Robert Lowenstein, Esquire s
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i LOW TEMPERATURE OVERPRESSURE
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EVENTS AT TURKEY POINT UNIT 4 i
Case Stucy Report i
Reactor Operations Analysis Branch Office for Analysis and' Evaluation of Operational 1)ata i
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t March 1954 l
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L Preparec :y:
wayne D. Lanning 1
1 NOTE:
This report cocuments results of stucy completec to cate by the Office for Analy-ts ano Evaluation of Coerational Data with regare to a l
particular operational situation.
The findings and recommencatient co not necessarily represent the position or recuirements of the responsitie program of fice nor the Nuclear Regulatory Commission.
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9404050445 840321 PDR ADOCK 050002S1 i
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- 1. 0 INTRODUCTION eefere 1979, 30 reportec incicents occurred in 'pressuri:ec water reactors (PWRs) where the pressure / temperature limits contained in the technical specifications for the reactor coolant system were exceeded.
Most of these events occurred during reactor startup or shutdown when the reactor coolant system was in a water solid condition, i.e., no steam or gas space in the pressurizer.
Over-pressure events primarily resulted from the loss of letcown flow with continued charging flow, inadvertent safety injection, or a heatup transient caused by starting a reactor coolant pump with the seconcary coolant system temperature higher than the primary temperature.
These events were causec by either ecuipment malfunction or operator error.
L;w-temperature overpresturi:ation (LTOP) was cesignated a generic issue because of the possibility of a vessel failing by the crittle fracture mechanism.
This failure mece may be a consecuence of a pressure transient after the vessel material toughness has been recuced cue to irradiation, effects (i.e., increase in nil-cuctility transition temperature) while a critical si:e flaw exists in the vessel wall.
NRC resolved the generic issue in 1979" by rec:mmending that PWR licensees implement procedures to reduce the potential for overpressure events and install equipment mocifications to mitigate such events.
Since that time, ten pressure transients have been reportec.
The two events at Turkey Point Unit : on Novemeer 28 and 29, 1981 exceecec the tecnnical s:ecification limit (415 psig below 355'F) by aDeut 700 anc 225 psi, respec-tively.
The t-o events were cesignated Abnormal Occurrences by the NRC (Ref. 1).
The other eignt reportec events were mitigated by the overpressure protection
' system.
Inese t.o overpressure events anc a significant number of events at Other P'='Es involving inopera:1e trains Of the Over;ressu-e er:tection system prom:tec AE;D : initiate an evaluation of operationai events with the focus
- -imari'y :n T rtey Peint, Ine overpressure protection system and the overpressure events at Turkey Point Unit a are cescrietc in Sections 2 and 3.
Sectien a contrains the analyses anc evaivatien of the two events, including utility managemeet's reaction to the even's.
Section 5 reviews the operational experience relatec to inoDerable trains of the overpressure protection system at other PwRs.
Section 6 evaluates the acecuacy Of existing LTOP technical specifications.
Section 7 ciscusses the 9eec f:r ::erating in a -ater solic cercitten.
Sect':n ! lists tne fine-ings anc concivsions. anc 5ection 9 contains the E;0 -e::mmencations :asec :n this :ase stucy.
i
- NUREG-0 2 e titlec, "teact:* Vessel Dressure Trarsient Pr:tection for Pres-L suri:ec Water Reacters.".as cuelishec in Se: tem er 1975 cocumentin; the ccm-
- letien f the generic activity.
LIOP mitigating systeas -ere installec in most 01 ants beginning in 1979.
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UNITED STATES
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'.,3 NUCLEAR REGULATORY COMMISSION y
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w AsmotoN. o. c. rosss j
.s.
March 11, 1987 Docket Hos. 50 250 and 50-251 i
i.
Mr. C. O. Woody, Group.Vice President l
Nuclear Energy Department i
Florida Power and Light Company i
Post Office Box 14000 t
' Juno Beach, Florida 33408
Dear Mr. Woody:
Subject:
Projected Values of Material Properties For Fracture Toughness Requirements'For Protection Against Pressurized Thermal Shock Events - Turkey. Point Plant, Units 3 and 4 j
Reference:
TAC Numbers 59992 and 59993 By letter dated January 23, 1986, and supplemented on June 5, and July 7,1986, i
you provided your response to the Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61 for the Turkey Point Plant, Units 3 and 4.
The staff, with the assistance of our contractor Brookhaven National Laboratory (BNL), have reviewed r
your submittals and performed confirmatory calculations, j
r Based on our review and confirmatory calculations, we have determined that the j
material properties of the reactor vessels beltline materials, the projected fluence at the. inner surface of the reactor vessels at the expiration date of l
at the expiration date of the licenses the licenses and the calculated RT (April 27, 2007) tobeacceptable,PTkhecalculatedRT both the licensee's f' -
and our confirmatory calculations, is well below the b e,ening criterion of 300'F for the limiting circumferential weld material at the expiration date of
)
1-the licenses and is therefore in conformance with the PTS Rule. The details l
of our evaluation and the basis for our conclusions are included in the enclosed Safety Evaluation.
l must be The PTS Rule requires that the projected assessment of the RT updatedwhenevercharigesincoreloadings,surveillancemeasubentsorother l
information (including changes in capacity factor) indtcate a significant E
This ensures that you will track the change in the projected values.
r tumulated fluence for the limiting beltline materials throughout the life of the plant to verify that your assumptions remain valid.
In this regard, we and comparison of request that you submit a re-evaluation of the RTPTS l
. 4,
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L Mr. C. O. Woody 7 t
the predicted value in any future Pressure-Temperature submittals which are submitted as required by 10 CFR 50, Appendix G, for each of the Turkey Point Units.
i This concludes our actions related to the above TAC numbers.
Sincerely, O a k % ject Manager Daniel G. Mcdonald, Senior Pro PWR Project Directorate #2 Oivision of PWR Licensing-A
Enclosures:
As stated cc w/ enclosures:
See next page i
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Mr. C. O. Woody Florida Power and Light Company Turkey Point Plant cc:
Harold F. Reis, Esquire Administrator Newman and Holtzinger, P.C.
Department of Environmental 1615 L Street, N.W.
Regulation Washington, DC 20036 Power Plant Siting Section State of Florida Mr. Jack Shreve 2600 Blair Stone Road Office of the Public Counsel Tallahassee, Florida 32301 Roem 4, Holland Building Tallahassee, Florida 32304 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Norman A. Coll, Esquire Suite 2900 Steel, Hector and Davis 101 Marietta Street 4000 Southeast Fimcial Atlanta, Georgia 30323 t
Center Miami, Florida 33131 2390 Martin H. Hodder, Esquire 1131 NE, 86th Street Mr. C. M. W ay, Vice President Miami, Florida 33138 Turkey Po1r.; Nuclear Plant Florida Power and Light Company P.O. Box 029100 Joette Lorion Miami, Florida 33102 7269 SW, 54 Avenue Miami, Florida 33143 I
Mr. M. R. Stierheim Mr. Chris J. Baker, Plant Manager County Manager of Metropolitan Turkey Point Nuclear Plant Dade County Florida Power and Light Company Miami, Florida 33130 P.O. Box 029100 Miami, Florida 33102 i
Resident inspector U.S. Nuclear Regulatory Commission Turkey Point Nuclear Generating Station Attorney General Post Office Box 57-1185 Department of Legal Affairs Miami, Florida 33257-1185 The Capitol Tallahassee, Florida 32304 Mr. Allan Schubert, Manager L
Office of Radiation Control I
Department of Health and I
Rehabilitative Services r
l 1317 Winewood Blvd.
Tallahassee, Florida 32301 Intergovernmental Coordination and Review I
Office of Planning & Budget Executive Office of the Governor l
The Capitol Building l
Tallahassee, Florida 32301 l
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NUCLEAR REGULATORY COMMISSION o
UNITE 0 STATES e
wassmotow, o. c.aosse
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$AFETY EVALUATION BY THE OFFICE OF NU REACTOR REGULATION REGARDING PROJECTED VALUES OF MATER 0,
i OPERTIES FOR FRACTURE TOUGHNESS REOUIREMENTS fE',,LROTECTION AGAINST PRESSURIZED TH FLORIDA POWER AND LIGHT COMPANY
^
TURKEY POINT PLANT, UNITS 3 AND 4 i
1.
Intecduction As required by'10 CFR 50.61, "Fract
- 4. 3 1
Against Pressurized Thermal Shock" (PTS Rure Toughness L
Feceral Racister on July 23 i
ule the licens)ee for each operating pre which was puolished in the water reactor "shall turit projec,ted valu
, 1985 of submittal'to the expiration date of the opsurfac es of RT zed y gbng(at the inner vessel must specify the bases for the=
values from the time erating license.
core loading patterns. _ This as' projection including' the assump c '
The assessment and must be updated whenever changes in core lo di
.essment must be submitted by January or'other information indicate a significant cha arding 23, 1986 a
ngs, surveillance meas By.' letters dated January nge in projected values."urements
-the Florida Power and Light Compan,y submitted i fand sup 23, 1986 the reactor pressure vessel, in compliproperties and the n ormation on t5e material 1
ance with the requirements of 10 CF 50.61 for the Turkey Point Plant, Units 3 l
'were projected to April and 4.
The RT j
27, 2007, which is th and fluence values Evaluation of The Material Aspects e expiratio$Tbate of both licensees.
II.
was identified to be intermediate-to-loTne controlling e standpoint of PTS susceptibility number 71249) for both unit 3 and unit 4 wer girth weld SA-1101 (weld wire heat it.e material properties of the controlli and chemistry factor were reported to be:g material and the associated n
margin Cu (copper content, %)
Utility Submittal Staff Evaluation 0.26 Ni (nickel content, %)
0.26 0.60 I (!n'
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)
0.60
+10
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Utility Submittal Staff Evaluation M (Margin, 'F) p a8 CF (Chemistry Tactor, 'F)-
j 166,8 The contro11 tog material has been properly identified i
for the copper and nickel contents and the' initial RT The justifications
. reference to a submitthi dated February staff an April 10, 1984, whibT are given by meet our criteria for FTS submittals.26, 1984 (S.A. Varga to J.W. William 5
50.61 of 10 CFR Part 50. consideration of the bases for these Assuming that the e,-Section shown above. correct, ' Equation 1 of the PTS rule governs, reported values of fluence and the chemistry factor is as III. Evaluation _of the Fluence Asoects Earl (a) y studies of the-PTS issue for the Turkey Point plants indicated the controlling beltline material is the intermediate t ferential weld SA-1101 and (b) a 1'9x reduction factor o-lower circum-be effected for both plants to prevent them from reachiof about 4.5 should screening criteria before April 2007 (i.e., the ex licenses).
ng the 10 CFR 50.61 l
r operating on the use of part-length absorber rods located on the assembli core flats.
sed the flux reduction measures and to evaluate the project es on the peak azimuthal fluence at the end of the current license on th of e estimate of the ferential welds.
e lower circum-The licensee's determination of the fast fiux at the lower is based on the 00T 4.3 discrete ordinates transport cod circumferential weld The calculations emoloy a nuclear data library based on the 47,0) geo e in (r BUGLE-80 (ENDF/8-IV) library, and an $s-P 3 angular decomposition. neutron group source is obtained from P0Q-7 generated pin-wis The neutron butions.
ower distri-source nn malization factor.
fwrenti41 weld is than given by:The fast (E > 1.0 MeV) flux at the lower circum-re neutron i
$ weld = $00T(r=PV inner surface, 0) P (z= weld elevation) 1.e., the DOT 4.3 (r,0) result is multiplied by the relativ elev6 tion of the welo (from a NODE-P calculation) to provide an es e axial power at the the three-dimensional fast flux at that location mate of The basic elements of the Brookhaven National Laboratory (BNL) approach for determining the fast flux at the peak wall location circumferential pressure vessel welds are summarized below:
, our contractor's, on the lower t
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Forward and/or adjoint fix
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00T-4.3 in contribution (s.0) and (r,z)ed source calculations are performe r
of selected assemblies and axial tones to thg flux; at the lower circumferential weld 1
~
e peak Azimuthal location).
eE
' s, near the core major ax>is (the1.0 MeV 2.
The 00T calculations employ a 16 neutr ENOF/B-IV based on 100 neutron group EPR libon(
decomposition.
F rary and an Sa*P 3.
with the 00T-4 Cycle specific source data provided b 3 angular i
distributions are accounted for viaassembly a ed sources are considered, and the neglect from an earlier study, ux. Only An exposure correction is applied a generic adjustment factor determinede d
4, dependent source spectrumeffect of plutonium on both
. 1 ormalization and the energy-resulting values of RTResults for the present and project d e
weld near the core majbg at the inner surface of the loend of license fast L
Units 3'and 4 axis are given in Tables 1 and 2 for 1
wer circumferential respec exposure (Case,s 1 vs. tively.
for estimating the three-dimension l 2 and 3 vs. 4), and the licensee vs ey Point ects of For Unit 3:
flux at the' limiting location. the BNL approaches a
EOL conditions (1) the exposure effect is worth 3.5% and 7%
present fluence, by ~2respectively; (2) the axial treatment between the licensee a% and the EOL fluence by ~10%; and (3) uni present and For Unit 4 nd BNL Case 1 results is <~3%.
the difference the cases w,ith no exposure correctithe exposure dependent resu have a smaller effect (<4%.
and the different axial treatments,w a s on and those from Case 1 show)an ~However,, comparison of the licens It is significant that 12% discrepancy (vs. <~3% for Unit 3) ee results are higher than those o,btained by the liceeven though the BNL result RT weNk,are still well below the NRC screening cr,iterion of 3 respectively.with end of license values of 271*F the resultant values (forCase 4) nsee ce Therefore, we conclude that the prop in.a RT PTS which meets the 10 CFR 50s61 criter osed flux reduction results IV.
Conclusion s acceptable.
Both the licensee's and our confirma' screenin licenses.g criteria for the limiting material at the expir ti and 4, respectively.The licensee has calculated a RTtory calcu e
a 271*F and 276*F for Units 3 and 4 Evaluation, the staff s on date of the y calculations are higher with a RTof this Safety weld material to April
, which is the expiration date of both li 27, 2007 of es.
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4 We therefore conclude that the Turkey Point Units circumferential~ welds. assemblies for the reduct on to the end of their e special wer throughout the life of the Turkey Point PlantIn order for t to the predicted value with future Pressure-Tempera e RT and comparison required by 10 CFR 50, Appendix G.
submIIkalswhichare Date:
Principal Contributors:
P. N. Randall 1..
l.ois 1
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RTpTS for Turkey Point Unit-3 Case Present End-of-License (2)
Fluenceil)
RTPTS IE)
Fluence (1)
RTpy3 BNL-FP&L Axial
_Tr eat men t 1.- Zero Exposure.
1.31 237
- 2. Expe;Jre Corrected 1.35 239 262 2.10 2.25 266
_BNL 3-D Synthesis se
- 3. 2ero Exposure 1.33 23B
- 4. Exposure Corrected 1.37 240 2.31 267 2.47 271 FP&L 1.27 236 2.15 263
' (1). Fluence (>1.0 MeV) x 10-18 i
n/cm2 L
(2)RTPTS from Eqn.1 of 10CFR 50.61 i
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Present' and' Projected E0L. Fluence (>1.0 MeV) and i^
RTPTS for. Turkey Point Unit-4 End-of-License III
.Present(l)
RT TS(2)
Fluence (1)
RTPTS Fluence P
Case BNL-FP8L' Axial.
- Tr eatment
- 1. 2ero Exposure 1.33 238
.2.40 269
- 2. Exposure Corrected.
1.39 240 2.60 274 BNL-3-D Synthesis
- 3. Zero Exposure 1,32 238 2.48 271
- 4. Exposure Corrected 1.39 240 2.70
- 276 FP&L l'.19 233 2.16 263 o
2 (1) Fluence (>1.0 MeV) x 10-18 n/cm (2) RTpis from Eqn.1 of 10CFR 50.61 5
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