ML20214C731
| ML20214C731 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/12/1987 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20214C732 | List: |
| References | |
| B12305, NUDOCS 8705210079 | |
| Download: ML20214C731 (7) | |
Text
- _ _ _ _ _ _ _ _ _ _ _
w General Offices
- Selden Street, Berlin. Connecticut unY n7aNc=="=
P.O. BOX 270 mes =ma maa cowa=
HARTFORD, CONNECTICUT 06141-0270 L
L J (("ygg,"g (203) 665-5000 May 12,1987 Docket No. 50-245 B12305 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C. 20555 Gentlemen:
Millstone Nuclear Power Station, Unit No. I Proposed Revision to Technical Specifications Pressure-Temperature Limit Curves Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its Operating License, No. DPR-21, by incorporating the changes identified in Attachment 1 into the Technical Specifications of Millstone Unit No.1.
The proposed changes revise the pressure-temperature limits and the maximum allowable heatup and cooldown rates for the reactor coolant system.
The Millstone Unit No.1 Safety Technical Specifications contain limitations on allowable reactor coolant system pressures and temperatures.
The current limitati ns are valid to a corrected value of 11.7 Effective Full Power Years 9
(EFPY).ti) The corrected limitation on EFPY was necessitated by the change in the predicted shift in fluence values. The change in these v in General Electric Company Report Number NEDC-30833(a ues was determined from the material contained in surveillance capsule number 2, wNch was removed at the end of fuel cycle 9.
The unit could achieve 11.7 El PY as soon as July 15, 1987, assuming continued plant operation.
Revised limits have been calculated in order to continue operation beyond 11.7 EFPY. These limits reflect the predicted radiation-induced embrittlement of the reactor vessel through 16 EFPY.
Pressure-temperature limits are required by 10 CFR 50 Appendix G, " Fracture Toughness Requirements," to provide adequate margins of safety during any condition of normal operation, (1) 3. F. Opeka letter to C. I. Grimes, " Pressure-Temperatt ce Operating Limit Curves," dated August 14,1986.
(2) General Electric Report NEDC-30833, " Millstone Nuclear Power Station,
[\\
Unit 1, Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis," T. A. Caine, December,1984.
g )
I B705210079 870512 gge,i J
C//tO/6 y4 33,0i)
DR ADDCK 0500
)
ej
s
+
. e
,U.S. Nuclear Regulatory Commission" 1B12305/Page 2:
.May 12,1987
' including anticipated operational _~ occurrences. These limits. depend upon thej metallurgical properties of the reactor; vessel materials.i ~ The vessel beltline ;
l' region material properties change over the lifetime. of ;the-vessel due to' the r effects of neutron irradiation.. The amount of neutron irradiation to which these i
materials are exposed determines ; the shif t =. in the material's 1 reference
' temperature for nil ductility transition (RTNDT) Property values. RTNDT=ls.the -
temperature'at.which materials exhibit ductile behavior. The shift in this value -
can be. predicted from the results of tests of reactor. vessel: surveillance.
' specimens and from the calculational methodology of Regulatory Guide 1.99.
The pressure-temperature limits must therefore be modified " periodically to1 l
reflect the vessel's reduced resistance to brittle fracture due to' irradiation. This will ensure that stresses in the vessel will be held within acceptable limits.
DESIGN BASES The original design ' bases for the plant included protection against brittle :
fracture.
These were described in FSAR Section 4.2.3 - and the Technical-Specification Basis Section 3.6..
The pressure-temperature ' limits. were -
t calculated in accordance with ASME III, Summer 1965 Addenda. Allowance for radiation embrittlement was included in accordance with guidance provided by the NSSS vendor.
The current operating limits are based on-calculational methods contained in!
ASME Code,Section III and _10 CFR Part 50, Appendix G, May,1983..- At the time that the Millstone Unit No.1 reactor vessel was fabricated (1965), the fracture toughness requirements were not as comprehensive as the current ASME.
Code Section III NB-2300 requirements. However, Paragraph III.A of 10 CFR 50 Appendix G states that an. approved method may be used to demonstrate equivalence of pre-1972 Code fracture toughness data with post-1972 Code.
requirements.
Toughness property correlations were derived :for the ~ vessel.
materials in order to use the available data to give a conservative estimate _ of RTNDT, compliant with the intent of 10 CFR 50 Appendix G criteria. These.
toughness correlations vary, depending upon the specific material analyzed, and -
were derived from the results of WRC Bulletin 217, " Properties of. Heavy Section Nuclear Reactor Steels," and from toughness data for other BWR reactors..The revised operating limits are based on calculational methods contained in ASME-Code Section III and the 1985 Edition of 10 CFR Part 50,6 Appendix G. The l
i operating limits are proposed for operation through -11.74 X 10 MWDth at which time the neutron fluence at the 1/4T (one-fourth the thickness of the vessel wall, measured from the inside) location in the reactor vessel wall will be about 4.3 X -
i; 1017 n/cm2.-
The most recently proposed method of predicting : radiation L
embrittlement, Draft Regulatory Guide 1.99, Revision 2, was used as the basis.-
DESIGN CALCULATION PARAMETERS.
U The Millstone Unit No.- I reactor vessel was manufactured by _ Combustion :
Engineering for General Electric. The vessel inner diameter is 226-1/16 inches,.
minimum wall thickness of 5% inches, with a' clad thickness ~of 1/8 inch. The chemical compositions for the plates in the beltline region, the weld metal from -
the longitudinal and girth seams joining the beltline plates and the recirculation -
and girth seams joining the. beltline plates and the recirculation inlet nozzles were determined from Table 3-1 of GE Report number NEDC-30833, as revised -
ig.,
,, 4p
+-
v gM ee--
yr --
g3 y-7 gy r
p-s-g 9
r y
w g
y g-y g
.a
.g-gg y
9km,_w, i
~
- 4..
v a
e(
UJi, Nuclear Regulatory Ccmmission'
~
a B12305/Page 3 c May"12,1987 -
1 "in the Errata:and Addenda, dated June ~ 1986.(3) ? An investigation o'f the chemical composition data contained in GE Report Number NEDC-30833 determined that:
the surveillance weld thought to be longitudinal seam weld W5214, was_actually-welded with the girth seam weld metal'34B009.
The recircuiation inlet nozzles were investigated to determine ~ if, due to their '
' proximity to the core, they were a concern.
GE provided :the following elevations and dimensions:
~
Elevations
-Bottom of Active Fuel:
211 inches Recirc. Inlet _ Centerline:
200 inches Recirc. Outlet Cente;line: 145 inches Dimensions Recirc. Inlet " Top":
'212-3/16 inches Recirc. Outlet " Top":
172-1/2 inches The peak surface fluence as listed in GE Report number NED'C-30833, page 4, is..
18n/cm2 or 32 EFPY, The percentage of relative flux at the location of -
1.6 X 10 f
the nozzle are as follows:
24% of peak at the top _
+
13% of. peak at the bend radius
~
- 3% of peak at the nozzle centerline-The recirculation nozzles were examined, but-the longitudinal weld #1-073 A/C (at 1/4T location) and the lower shell plate #G-2002-3 (at the 3/4T ' location) were determined to be the limiting locations (see Table 1, Attachment 2).. The-chemical content, surface fluence and-RTNDT. values. for potentially. limiting _
locations at 16 EFPY are presented in Table 1.
The surface fluence for the limiting locations is listed in.GE' Report Number NEDC-30833 as 1.6 X 1018n/cm2 at 32 EFPY.' This1value includes a 25%
~
increase to account for uncertainty in the. flux' wire test results. However, the; embrittlement prediction of Draft Regulatory. Guide 1.99, Revision 2, was used, which allows two standard deviations to account 'for error (480F).H Since this
'i error allowance was made in -the-calculation, the 25% increase on fluence, u
proposed by GE, was deducted.
The attenuation ratio'ior the flux was estimated using the' "dpa equivalent" -
exponential decay model with an exponent of.24x, where x is the depth in (3)
Errata and Addenda to GE Report NEDC-30833, " Millstone Nuclear Power Station, Unit No.1, RPV : Surveillance - Materials-Testing ! and Fracture-Toughness - Analysis," - to this letter. ~
dated - June, 1986, provided ' as u
U.S. Nucl:ar Rcgulatory Ccmmissicn -
B12305/Page 4 May 12,1987 inches. This yields a 1/4T thickness fluence of 4.3 X 1017 and a' 3/4T thickness -
fluence of 2.2 X 1017 n/cm2 (E hl MeV) at 16 EFPY for the limiting weld and plate.
The limiting RTNDT at 16 EFPY for the weld and plate as determined using-Regulatory Guide 1.99, Rev. 2 are as follows:
1/4T, RTNDT( F) 3/4, RTNDT( F)
G-2002-5 95.65 84.31 1-073A/C 104.01 82.36 Pressure-temperature limits for heatup, cooldown, isothermal operation and hydrostatic testing were developed in accordance with 10 CFR 50 Appendix G-and ASME III, Appendix G. The most limiting case of inner and outer diameter flaws were used._ Thermal calculations were performed using ABAQUS, a general purpose finite element computer code. The heatup transient was calculated starting at 700F, heating the coolant at a rate of 1000F per hour to 5500F The cooldown was calculated by. starting at 5450F and cooling at a rate of 1000F per hour to 700F. In both cases, the film coefficient was infinite and a 1/8 inch stainles: steel clad was included.
The indicated pressure was corrected to allow for ebvation head differences'and measurement error. The elevation head from the recirculation inlet nozzle to the vessel head top inside elevation is 20.57 psi at room temperature.
Additionally, a measurement error of 12.5 psig was included for all temperatures. A measurement error for the temperature instruments of 4.310F was also included.
The design pressure for this vessel is 1250 psig and it normally operates at 1000 psig.
POSTULATED FAILURES The intent of the operating pressure-temperature limits is to prevent brittle fracture of the reactor coolant pressure boundary. If a failure. were to occur, it would present an unanalyzed accident. The possible-modes of this postulated fracture would be via operation in violation of the calculated limits, use of incorrect bases used in determining the limits, or a miscalculation of the limits.
The following safeguards are employed to ensure that such a failure is not a credible event:
Miscalculation - The Northeast Utilities Quality Assurance Calculation o
Procedure is used to avoid calculational errors. This procedure requires a full independent check and supervisory approval. To further reduce the chance of error, the calculational technique has been codified (Draft Regulatory Guide 1.99, Revision 2). Gross miscalculation error' is thus precluded.
o Incorrect Bases - There are key bases to this calculation which must be correct to ensure the validity of the resulting operational limits. These
y y,.. - - ~
L C
N
]
'7 1,;
v...
7 U.S. Nuclear Regulatory Ccmmissisn.j g
.j iz
^
~ LB12305/ Pag 2 5.-
-l
. May l2,'1987; t
- 1
- bases include the' bounding flaw size an'd the$ bounding matdrial frEture toughness properties.
EThe.. bounding flaw size,is~ ~one-fourth; the. thickness of the vessel wall, measured from the inner diameter of the vessel. ' This assumed flaw size !
E w
is assured. to be bounding byf actual preservice= inspection' measurement verification and design margins against in-service cracking.
Thel ounding material fracture toughness properties are calculated as b
~
discussed in the current design bases summary of: this evaluation.'These
. calculations are comp'ared with surveillance icapsule: results.tSuch a capsule, which : contained ; material. test. coupons, was : removed i from Millstone Unit No.5 1 during the 1984 refueling outage. The caps,ule was -
~
analyzed and the shift in material properties was greater than predicted.
This revision reflects the shif t 'and the limits have.been adjusted to' be valid to 16 EFPY. The-calculational prediction and the surveillancej capsule verification of the prediction assure that bounding propertie's are -
used.
As a result, these key calculational bases are assured to be bounding.
..nu.
o Violation of Limits - Limits which are correctly, calculated with bases which are properly determined will. assure that brittle-fracture of the-reactor vessel will ~not occur, provided ~the, limits are not'. violated.'
Violation of the limits is precluded by the following safeguards:
The calculation performed to determine the limits -employs)a safety factor of two (2) on the pressure v~alues.1 The assumptions -
upon which the calculation is based are 'also conservative, and provide an even greater overall_ safety factor. Consequently, even a pressure value more than two times the limit would notl be expected to propagate a postulated bounding one-fourth thickness - -
crack.
Extremely large temperature change rates are not expected to be.
capable of. producing a through.wallu fracturej in a single occurrence. Thermal stresses which result are relieved. by the' fracture itself.
The combination of these protective measures assure that those' limit violations which can occur would not produce failures from brittle fracture.1 In summary,,
due to the safeguards noted above, a postulated fracture which would result in an unanalyzed accident is not credible.
NNECO has reviewed the attached proposed changes pursuantjto 10CFR50.59 t
and has determined that they do not constitute an.unreviewed safety question.~.
The probability of occurrence or the consequences of an accident'or malfunction c of equipment important to ' safety' (i.e., safety-related) previously.' evaluated in' w
the final safety analysis report have not been increased. The possibility for an-
. accident or malfunction of a different type than any evaluated previously in the.
~
final safety analysis report has not been created. There has n_ot been a reductioni E in the margin of safety.as defined in the basis for any; Technical. Specification.
The proposed Technical Specification changes ensure that the reactor vessel is maintained within its original design tolerances for all unit operations..
e
~
.u f
A
-t
U.S. Nucliar Rtgulatory Crmmission B12305/Page 6 May 12,1987 NNECO has reviewed the proposed changes, in accordance with 10CFR50.92, and has concluded that they do not involve a significant hazards consideration in that these changes would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated. Since the design basis safety factors have been maintained, there.is no such increase in accident or malfunction probabilities. Since this change is an update of the existing operational limits, reflecting increased vessel neutron-induced embrittlement,'
accident or malfunction consequences are not adversely affected.
2.
Create the possibility of a new or different kind of accident from any previously analyzed. Again, since this change is an update of existing limits without any hardware modifications, no new accidents or malfunctions are created.
3.
Involve a significant reduction in a margin of safety. As previously stated, this change will not adversely affect the current margins of safety.
Safety margins as defined in the bases of the - Technical Specifications are maintained.
The Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (51 FR 7751, March 6,1986). The changes proposed herein most closely resemble example' (ii), a change -that -
constitutes an additional limitation, restriction or control not presently included in the technical specifications. The proposed heat-up and cool-down curves are more restrictive than the existing curves. The basis of the new curves is the same as the basis of the current curves, merely updated to reflect an interval of time later in service life of the reactor pressure vessel.
The Millstone Unit No. I Nuclear Review Board has reviewed and approved the attached proposed revision and has concurred with the above determinations.
In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment.
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment request is the application fee of $150.00.
These proposed changes are not required to support plant start-up from the 1987 refueling outage since the existing heat-up and cool-down curves will remain valid for a short int 9rval into Cycle 12 operation. However, NNECO respectfully
)
requests that the Staff target issuance of the requested licensee amendment by the end of the 1987 refueling outage.
Very truly yours, l
\\
NORTHEAST NUCLEAR ENERGY COMPANY l
AW E. M ffoczka f7 Senior)Vice President l
J m
U.S. Nucle:r Regulat:ry Ccmmission '
B12305/Pagn 7 May 12,1987 cc: W. T. Russell, Region I Administrator
- 3. 3. Shea, Project Manager, Millstone Unit No.1 -
T. Rebelowski, Resident Inspector, Millstone Unit No. I Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protec+ ion Hartford, Connecticut 06116 STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD
)
Then personally appeared before me E. 3. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized - to execute and file the foregoing.
information in the name and on behalf of the Licensee herein and that the -
statements contained in said information are true and correct to the best of his
-knowledge and belief.
Ndtary Pubh
~ dif/NW 1st My Commission Expires March 31, ISH 1
N
-4
+
v