ML18010B049
| ML18010B049 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/26/1993 |
| From: | Starkey R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18010B050 | List: |
| References | |
| NLS-93-059, NLS-93-59, NUDOCS 9303090314 | |
| Download: ML18010B049 (28) | |
Text
ACCELERA.
DOCUMENT DIST UTION SYSTEM
'EGULA.
eY INFORMATION DISTRIBUTIG.
SYSTEM (RIDS)=
ACCESSION NBR:9303090314 DOC.DATE: 93/02/26 NOTARIZED:'YES DOCKET FAC:.L:50-400 Shearon Harris Nuc'ear Power Plant, Unit 1, Carolina 05000400 AUTH.NAME AUTHOR AFFILIATION STARKEY,R.B.
Carolina Power a Light Co.
RECIP.NAME 'ECIPIENT AFFIlIATION Document Control Branch (Document Control Desk)
SUBJECT:
Application for amend to License NPF-63,revising Tech Spec 3.4.9 by replacing current five yr heatup 6 cooldown-limitations w/revised limitations based on predicted
- reactor, vessel neutron exposure at ll EFPY of operation.
DISTRIBUTION CODE.'OOID COPIES RECEIVED:LTR g ENCL I
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NOTES:Application for permit renewal filed.
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CML Carolina Power & Light Company P.O. Box 1551 ~ Raleigh, N.C. 27602 FEB 36 1883 R. B. STARKEY,JR.
Vice President Nuctear SeNtces Oepartment SERIAL:
NLS-93-059 10 CFR 50.90 United States Nuclear Regulatory Commission ATTENTION:
Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power
& Light Company (CP&L) hereby requests a revision to the Technical Specifications for the Shearon Harris Nuclear Power Plant (SHNPP).
This Technical Specification change revises Specification 3.4.9 by replacing the current five year heatup and cooldown limitations with revised limitations based on the predicted reactor vessel neutron exposure at 11 Effective Full Power Years of operation.
The revisions affect the Reactor Coolant System pressure-temperature limitations and the effective temperature ranges of the Reactor Coolant System heatup and cooldown rates.
In addition, the proposed amendment includes other related changes as identified in the Enclosure l.
Enclosure 1 provides a detailed description of the proposed changes and the basis for the changes.
Enclosure 2 details, in accordance with 10 CFR 50.91(a),
the basis for the Company's determination that the proposed changes do not involve a significant hazards consideration.
Enclosure 3 provides an environmental evaluation which demonstrates that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
no environmental assessment needs to be prepared in connection with the issuance of the amendment.
Enclosure 4 provides page change instructions for incorporating the proposed revisions.
Enclosure 5 provides the proposed Technical Specification pages.
9303090314 930226 PDR ADOCK 05000400 P
PDR (0
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In accordance with 10 CFR 50,91(b),
CP&L is providing the State of North Carolina with a copy of the proposed license amendment.
In order to allow time for procedure revision and orderly incorporation into copies of the Technical Specifications, CP&L requests that the proposed amendments, once approved by the NRC, be issued such that implementation will occur within 60 days of issuance of the amendment.
Please refer any questions regarding this submittal to Mr. R.
W. Prunty at (919) 546-7318.
Yours very truly, R. B. Starkey SDC/sdc
Enclosures:
1.
Basis for Change Request 2.
10 CFR 50.92 Evaluation 3.
Environmental Considerations
..4.
Page Change Instructions 5.
Technical Specification Pages R. B. Starkey, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are
- officers, employees, contractors, and agents of Carolina Power
& Light Company.
My commission expires:
Q-'f. 9Q cc:
Mr. Dayne H. Brown Mr. S.
DE Ebneter Mr. N. B. Le Mr. J.
E. Tedrow Notary (Seal) llllllllll
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ENCLOSURE TO SERIAL:
NLS-93-059 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS BASIS FOR CHANGE RE UEST
~Back round General Design Criterion 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," of Appendix A to Title 10 of the Code of Federal Regulations, Part 50, requires that the reactor coolant boundary be designed with sufficient margin to assure that when stressed under operating, maintenance,
- testing, and postulated accident conditions:
(1) the boundary behaves in a nonbrittle
- manner, and (2) the probability of rapidly propagating fracture is minimized.
Appendix G to 10 CFR 50, "Fracture Toughness Requirements,"
describes specific requirements for fracture toughness and reactor vessel operation to meet the 10 CFR 50.60 criteria regarding prevention of brittle fracture.
In addition, Appendix G requires the effects of changes in the fracture toughness of reactor vessel materials caused by neutron radiation throughout the service life of nuclear reactors to be considered in the limits on operation.
Regulatory Guide 1.99 contains procedures for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.
Paragraph V.B of Appendix G states that the reactor may continue to be operated only for that service period within which the requirements of Section IV of Appendix G are satisfied using the adjusted reference temperature and the predicted value of the upper shelf energy at the end of the service period to account for the effects of radiation on the fracture toughness of beltline materials.
Paragraph V.A requires that the effects of neutron irradiation on the reference temperature and upper shelf energy of beltline materials, including welds, be predicted from the results of studies in addition to the results of the Capsule Surveillance Program of Appendix H to 10 CFR 50.
In accordance with 10 CFR 50.36(c)(2), limiting conditions for operation are to be included in a plants'echnical Specifications (TS).
Technical Specifications 3.4.9.1 and 3.4.9.2, "REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS," provide Reactor Coolant Syst: em (RCS) pressure-temperature limits to protect the reactor pressure vessel from brittle fracture by clearly separating the region of normal operations from the region where the reactor vessel may be subject to brittle fracture.
The heatup and cooldown rates of Specifications 3.4.9.1 and 3.4.9.2, and the LTOP setpoints in Specification 3.4.9.4 are designed to ensure that t'e 10 CFR 50 Appendix G pressure-temperature limits for the RCS are not exceeded during any condition of normal operation including anticipated operational occurrences and system hydrostatic tests.
The current RCS pressure-temperature limitations for SHNPP were developed in accordance with 10 CFR 50 Appendix G criteria and the calculative procedure
ENCLOSURE TO SERIAL:
NLS-93-059 for determining the adjusted reference temperature in Position 1.1 of Regulatory Guide 1.99, Revision 2, for a predicted reactor vessel neutron irradiation equivalent to five Effective Full Power Years (EFPY) of operation.
The five EFPY curves were approved by NRC on December 23,1990 as Amendment 23 to the Operating License.
The Shearon Harris Nuclear Power Plant is expected to complete five EFPY in August 1993.
For this reason, new pressure temperature limits are needed for continued operation of the plant.
In compliance with t'e 10 CFR 50 Appendix H Reactor Vessel Material Surveillance
- Program, a surveillance capsule was removed at SHNPP at the end of Cycle 1 and again at the end of Cycle 3 for testing and examination.
The capsules contained samples of the limiting reactor vessel wall plate material.
Results from fracture toughness tests of the capsules were used to calculate the actual Shift in Nil Ductility Reference Temperature (dRTgpy) for the limiting plate material.
Capsule dosimeter results were used, in part, to determine the actual neutron fluence experienced by the reactor vessel wall.
Utilizing USNRC Regulatory Guide 1.99 Revision 2, Position 2.1, a new chemistry factor was obtained using the actual AT~~ and fluence from the surveillance program results.
The new chemistry factor and estimated fluence at 11 EFPY were used to predict the Adjusted Reference Temperature (ART+>z) for the current limiting plate (B4197-2) at 11 EFPY.
- However, 10 CFR 50 Appendix G paragraph V.B also requires that the ART~> be calculated for all beltline materials and that the highest ART~< be used to verify that the fracture toughness requirements are satisfied.
The ART~q for other beltline materials were recalculated for ll EFPY service life based on the calculative method of Reg.
Guide 1.99, Revision 2, Position 1.1, since no surveillance capsules with these materials were available.
Using this approach resulted in the determination of a new limiting plate material, A9153-1.
Based on the new limiting plate A9153-1 at ll EFPY, revised pressure-temperature limits were established for normal operation and for In-Service Leak and Hydrotests (ISLH).
As a result, of the beneficial effects of the Reactor Vessel Capsule Surveillance Program and the implementation of Regulatory Guide 1.99, Revision 2, Position 2.1, the revised pressure-temperature limits for any particular heatup or cooldown rate are less restrictive for 11 EFPY than for the existing 5 EFPY.
That is, at any given temperature and heatup or cooldown r'ate, the allowable pressure limit has increased.
Pro osed Chan e
This Technical Specification change revises Specification 3.4.9 by replacing the current five EFPY heatup and cooldown limitations with revised limitations based on the predicted reactor vessel neutron exposure at ll EFPY of operation.
El-2
ENCLOSURE TO SERIAL:
NLS-93-059 The revisions affect the RCS pressure-temperature limitations and the effective ranges of the RCS heatup and cooldown rates.
In addition, the proposed amendment includes other associated changes such as:
identifies a new controlling reactor vessel beltline material, relocates allowance for instrumentation uncertainties from the curves of Figures 3.4-2, 3.4-3, and 3.4-4 to plant implementation documents, extends the 10'F cooldown rate below 115'F, revises the Hydro and Leak Test curves and their presentation, rewords the implementation footnotes for the heatup and cooldown rates in Table 4.4-6 to improve clarity, revises the BASES to reflect these changes as well as incorporation of a revised Figure B 3/4.4-1 on Fast Neutron Fluence.
Basis Using the RG 1.99, Revision 2, Positions 1.1, 2.1 and 10 CFR 50 Appendix G
- criteria, a new Adjusted Nil Ductility Reference Temperature (ART~q~) and limiting pressure-temperature curves were prepared for the projected reactor vessel exposure at ll Effective Full Power Years (EFPY) of operation.
The new 11 EFPY curves impose less restrictive limits on plant operations than do the previous five EFPY curves.
As a result, the effective temperature ranges for the heatup and cooldown rates were adjusted.
The revised ranges, in conjunction with the current rates and LTOP setpoints, were chosen to: 1) ensure that the Appendix G pressure-temperature curves are not challenged given a limiting mass or heat input to the RCS~ during normal operations, anticipated occurrences and system hydrostatic testing, and 2) ensure that operational flexibilityis maintained.
Discussion Surveillance Capsule Results:
In compliance with 10 CFR 50 Appendix H, which describes the requirements for a Reactor Vessel Material Surveillance Program and invokes the requirements of ASTM E 185-82, two surveillance capsules were removed at SHNPP at the end of Cycles 1 and 3 for testing and examination.
The results of the tests and a report were submitted to the NRC (References 1 and 2).
In particular, the actual shift in nil ductility reference temperature (dRT~z) was calculated based on the fracture toughness
- tests, and the actual neutron fluence experienced by the reactor vessel wall was established based in part on the capsule dosimeter results.
Limiting mass input - inadvertent startup of one charging/safety injection pump Limiting heat input - inadvertent startup of one reactor coolant pump while the steam generator secondary side is 50'F higher than the primary side
ENCLOSURE TO SERIAL'LS-93-059 As required by 10 CFR 50 Appendix G, paragraph V.B, the predicted value of the adjusted nil ductility reference temperature (ART~+) has been calculated for 11 EFPY based on several factors.
As noted and described in CP&L's previous submittal to the NRC (Reference 2), the 'ow leakage" core concept implemented in Fuel Cycle 2 will continue to be utilized for future fuel cycles.
Therefore, a revised fast neutron fluence (E >
1 Mev) versus EFPY curve is proposed to replace Figure B 3/4.4-1 of TS BASES -3/4.4.9.
Based on this, the neutron fluence that will be experienced by the reactor vessel beltline materials has been predicted for future fuel cycles (including up to 11 EFPY).
As a result of testing and examination of the Surveillance Capsules "U"
and 'V" removed from the reactor vessel at the end of Cycles 1 and 3
(approximately 1 EFPY and 3 EFPY respectively),
a revised withdrawal schedule for future capsules has been proposed as discussed in the CPGL submittal (Reference
- 2) of the Reactor Vessel Material Surveillance Report to the NRC.
This submittal does not affect the coupon withdrawal schedule proposed in Reference 2.
Utilizing USNRC Regulatory Guide 1.99 Revision 2, Position 2.1, (Ref.
3) a new chemistry factor (42 versus 65 originally) was obtained using the actual AT~~ and fluence from the surveillance program results.
This new chemistry factor, when combined with the predicted fluence for ll EFPY using the methodology of RG 1.99 Revision 2, Position 1.1, produced an adjusted reference temperature (ART~<) for the reactor vessel plate material (B4197-2), i.e.,
the existing controlling (limiting) material, of 147'F at the T/4 location and 133'F at 3T/4 (T reactor vessel wall thickness).
This is a significant improvement over 185'F Q T/4 and 169'F Q 3T/4 which is the ART>D> that would have been obtained using only the conservative calculation method of Position 1.1 of RG 1.99 Revision 2 without the benefits of the Surveillance Program results.
Controlling Material:
10 CFR 50 Appendix G paragraph V.B also requires that, the ARTzDz be calculated for all beltline materials and that the highest ART~> be used to verify that the fracture toughness requirements are satisfied.
The surveillance capsules
- though, only contain the most limiting
- Plate, B4197-2, (worst case initial RT~z and worst case copper and nickel contents, i.e., chemistry factor) before the vessel was irradiated, and some beltline weld materials.
Other reactor vessel beltline materials subject to high irradiation are not in the surveillance capsules.
As a result, the ARTqDq for other beltline materials were recalculated for 11 EFPY service life based on RG 1.99 Revision 2, Position 1.1.
Using this approach, with a chemistry factor of 58, the highest ART~~
at 11 EFPY was 148'F Q T/4 and 133'F Q 3T/4 for Plate A9153-1.
Therefore, there is a slight increase in ART~> when comparing Plate A9153-1 with the previous controlling Plate Material B4197-2.
In accordance with 10 CFR 50 Appendix G paragraph V.B, the revised
ENCLOSURE TO SERIAL:
NLS-93-059 pressure-temperature limits of operation for ll EFPY are therefore to be based on an ART~> of,148'F Q T/4 and 133'F Q 3T/4 for a new controlling material Plate, A9153-1.
Although the new Controlling Material A9153-1 is applicable for 11 EFPY, the actual switchover of controlling materials from B4197-2 to A9153-1 is estimated to occur at 8 or 9 EFPY operation.
Prior to this service life, plate Material B4197-2 continues to control.
Using the ART~> at ll EFPY for Material A9153-1 is more restrictive than the ART~~ for either material between five and 11 EFPY using t'e new fluence curve of BASES Figure B 3/4.4-1.
Therefore, its use is conservative and bounding.
The results of the fracture toughness tests including reactor vessel end of life (EOL) Upper Shelf Energy (USE) for all beltline materials were submitted to the USNRC in Reference 2.
The EOL USE is predicted to be 55 ft. lbs. for Plate B4197-2 and 60 ft. lb'or Plate A9153-1 at the T/4 location.
All other beltline materials, including welds, are postulated to have higher EOL USE than for B4197-2 or A9153-1.
The predicted EOL USE for each of the beltline materials is greater than the 50 ft. lbs.
minimum level as required by 10 CFR 50 Appendix G paragraph IV.A (1).
Material Properties:
'The reactor vessel toughness data of TS BASES Table B 3/4.4-1 has also been updated.
The phosphorous composition has been replaced by that of nickel for all beltline materials since the nickel and copper compositions are now used to determine the appropriate chemistry factor when utilizing Regulatory Guide 1.99 Revision 2 for determining ART~>.
- Also, as previously submitted to the NRC (Reference 2), recently obtained additional data requires the B4197-2 plate initial RT~z be increased from 86'F to 91'F and the initial USE be decreased from 74 ft-lbs to 71 ft-lbs.
The higher initial RT~q was utilized in determining the 11 EFPY ART>D> of 147'F.
Even though the initial RT~z increased, it is no longer the limiting material since Plate A9153-1 now controls.
Pressure-Temperature Limits:
Utilizing a predicted ART~< of 148 F/133'F for plate A9153-1 at 11 EFPY, revised pressure-temperature limits were established for normal operation and for In-Service Leak and Hydrotests (ISLH).
These new limits were established following the methods of analysis and the required margins of Appendix G of the ASME Code in accordance with 10 CFR 50 Appendix G paragraph IV.A.2.
The specific procedures utilized were those described in Topical Report BAW-10046A Revision 2 (Reference
- 4) and accepted by the USNRC.
As a result of the beneficial effects of the Reactor Vessel Capsule Surveillance Program and the implementation of Regulatory Guide 1.99, Revision 2, Position 2.1, the revised pressure
- temperature limits for any particular heatup or cooldown rate are less restrictive for 11 EFPY
ENCLOSURE TO SERIAL:
NLS-93-059 than for the existing 5 EFPY.
That is, at any given temperature and heatup or cooldown rate, the allowable pressure limit has increased.
(i)
ISLH The ISLH pressure temperature limits are less stringent than the normal operational limits.
This is because a reduced safety margin is allowed by ASME Appendix G criteria and documented in the TS BASES 3/4.4.9 This reduction in margin is allowed partly because reactor vessel thermal stresses will be minimal if the temperature rate change is less than 10'F/hr.
and partly because no fuel should be in the vessel when the consequences of failure are less
- severe, as explained in the ASME Appendix G Source Input
- Document, WRC Bulletin 175 (Reference 5).
This is the reason for existing TS 3.4.9.1(c) and 3.4.9.2(c) which limit ISLH temperature rate changes to less than 10'F/hr.
The lower ISLH temperature limit is required to be 60'F greater than ART~~ for the controlling beltline material per the criteria of 10 CFR 50 Appendix G paragraph IV.A (4)
~
Based on the new Controlling Plate A9153-1 ART~> of 148'F at T/4 wall location and a thermal lag of approximately 33'F, the minimum temperature would have to be at least 241'F.
This criteria will be satisfied since the reactor vessel temperature also has to meet the limiting ISLH heatup curve minimum temperature of approximately 247'F for a pressure of only 1800 psig.
Therefore, the requirements of paragraph IV.A (4) of 10 CFR 50 Appendix G will be met.
The minimum ISLH temperature determined by the pressure-temperature limits for a hydrotest pressure of 2500 psia (2485 psig),
the reactor vessel design
- pressure, (or 10 percent above nominal operating pressure),
is 280 F at 10'F/hr.
ISLH cooldown rate of 10'F/hr. is also provided in case the reactor vessel temperature fluctuates up or down during the ISLH.
If fuel is in the reactor vessel during ISLH, then the normal heatup and cooldown pressure-temperature limits apply as required by paragraph IV.A (5) of 10 CFR 50 Appendix G.
(ii)
Core Criticality Limits:
Paragraph IV.A (3) of 10CFR50 Appendix G also requires a minimum temperature for core criticality to be at least above that for ISLH (280'F) and 40'F above the normal pressure-temperature operational limits, (approximately 345'F for 100'F/hr.
heatup at 2235 psig).
Technical Specification 3.1.1.4 requires, for core criticality, the lowest operating loop average temperature be at least 551'F.
Since this temperature is well above that required by the criteria of 10CFR50 Appendix G paragraph IV.A (3) (345'F plus 40'F or 385'F total),
the, core criticality limit lines have been omitted from TS Figure 3.4-3 (Reactor Coolant System Heatup Limitations) as previously accepted by the NRC in TS Amendment 19, (Reference 9).
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~
4i
ENCLOSURE TO SERIAL:
NLS-93-059 (iii) Closure Flange Limits:
In addition to beltline materials, Appendix G to 10 CFR 50 also imposes pressure-temperature limits based on the nil ductility reference temperature for the reactor closure flange materials.
Paragraph IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of the pre-service system hydrotest pressure (621 psig),
the temperature of the closure flange regions highly stressed by bolt pre-load must exceed the nil ductility reference temperature of the material in those regions by at least 120'F for normal operation and by 90'F for hydrostatic pressure tests and leak tests's discussed in TS BASES 3/4.4.9, the 5 EFPY pressure-temperature limitations were more limiting than the criteria of Appendix G
paragraph IV.A.2.
This was because the ART~~ for the beltline controlling material was originally predicted based on conservative methods and the fact that the closure flange region materials had a low initial RT~z and were not subject to high neutron irradiation and hRT~z compared to beltline materials.
However, since revised fluence curves are proposed (due to low leakage core concept),
and revised chemistry factors are used (based on actual surveillance data and the use of RG 1.99, Revision 2, Position 2.1), the proposed beltline material pressure-temperature limits for ll EFPY are less restrictive.
Therefore, a new comparison has to be made to ensure the limits of Appendix G paragraph IV.A.2 are not exceeded.
As discussed in TS BASES 3/4.4.9, the minimum temperature of the head closure flange and vessel flange region is 120'F because the limiting flange material RT~z is O'. If thermal lag were considered, then the minimum fluid temperature would need to be a few degrees higher.
The revised Appendix G 10'F/hr. heatup pressure-temperature curve for the vessel beltline materials at 621 psig has a minimum fluid temperature of 120'F also.
But when instrumentation uncertainties are considered, the 10'F/hr. heatup pressure-temperature limits would impose a temperature much higher than that required for the closure flange region even when considering thermal lag.
Similarly at 50'F/hr.
(the maximum normal heatup rate below 350'F),
the fluid temperature corresponding to a vessel flange temperature of 120'F would need to be approximately 174'F when considering conservative thermal lag effects.
At this 50'F/hr. heatup rate, the Appendix G
pressure-temperature limits for the vessel beltline materials at 621 psig would require a minimum fluid temperature of 170'F.
- However, the lower temperature limi.t allowed by proposed TS Table 4.4-6 for the 50'F/HR heatup rate is 175'F.
When instrumentation uncertainties are considered, the required minimum fluid temperature is much higher than the 174'F imposed by the flange materials.
Therefore, the closure head flange and vessel flange material minimum temperature is bounded by the Appendix G
l 1
l k
s
ENCLOSURE TO SERIAL:
NLS-93-059 pressure-temperature limits for the beltline materials when instrumentation uncertainties are considered.
Plant cooldown is not impacted because the beltline materials and flange materials are always hotter than the fluid temperature by the amount of thermal lag.
At 120'F fluid temperature, the 10'F/hr.
cooldown rate has a pressure limit of 608 psig and therefore bounds the flange material maximum pressure requirements.
Heatup and Cooldown Rates:
As discussed
- above, based upon a new controlling reactor vessel beltline material and an adjusted nil ductility reference temperature (ART~>)
for a new service life of 11 Effective Full Power Years (EFPY), revised pressure temperature limits were established and Figures 3.4-2 and 3.4-3 were redrawn to reflect the new limits.
In revising these
- curves, the rates of cooldown and heatup, in general were kept the same as the existing rates for 5 EFPY to minimize any impact on procedures and operator retraining.
- However, the effective temperature ranges for the established rates were revised as identified in proposed TS Table 4.4-6.
These temperature ranges were revised so that the reactor vessel Low Temperature Overpressure Protection System (LTOPS) can still provide the necessary reactor vessel protection for the design basis events and for the setpoints previously established.
The footnote to TS Table 4.4-6 regarding the location in the RCS from which to take temperature readings to use for heatup and cooldown rates and temperature ranges when the RCPs are not in operation was revised to remove ambiguity.
The footnote was added in TS Amendment 19.
The intent of the footnote remains the same.
The proposed rewording eliminates any reference to the use of lowest RCS cold leg value when the RCPs are not in operation.
In this manner all temperature
- readings, when the RCPs are not in operation, will utilize the RHR heat exchanger outlet temperature for both selecting the applicable heatup or cooldown
- rate, and for determining compliance (monitoring temperature in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period) with the selected rate.
This method ensures conservative operation of the plant.
RCS cold leg values will only be used when an RCP is in operation.
Low Temperature Overpressure Protection System (LTOPS):
The basic analysis methodology and design basis considerations described in TS BASES 3/4.4.9 remain unchanged.
That is, the WOG methodology (Reference
- 6) accepted for use at SHNPP (References 7 and 8) is utilized with consideration for the mass input from a safety injection pump start or the heat input from the start of a RCP with a differential temperature of 50'F or less between the secondary and primary systems.
The design basis considers the most limiting single failure (of one pressurizer PORV to open) for any reason.
The LTOP setpoint values as noted on TS Figure 3.4-4 and 'its enable
J t
k
ENCLOSURE TO SERIAL'LS-93-059 (arming) temperature have not been revised.
This is because the ARTSD~
at 11 EFPY for Controlling Plate A9153-1 is less than that currently utilized in the TS pressure-temperature curves for 5 EFPY.
The enable (arming) temperature for LTOPs is retained at its conservative value using the method of BTP RSB 5-2, Revision 1, paragraph B.2 for SRP Section 5.2.2 Revision 2.
The LTOPS design bases include consideration of the mass input from a Safety Injection (SI) pump start into a water solid Reactor Coolant System (RCS).
The existing method assumes the run out flow rate of an SI pump into a depressurized RGB The existing assumption is overly conservative as the RCS is water solid and normally pressurized during startup and shutdown in order to run the Reactor Coolant Pumps (RCPs).
As a result, any SI flow into the RCS would not be at the run-out flow but at reduced flowrate consistent with the RCS backpressure.
Therefore, the assumption for the mass input case has been revised to reduce this conservatism by considering SI pump operation in the SIS mode, but providing flow against an RCS backpressure consistent with the LTOP setpoint with due consideration for instrumentation uncertainties.
In contrast to the benefits of the above issues, some items adversely affected the LTOP analysis.
Namely the dynamic pressure drop between the LTOP pressure sensor location and the reactor vessel beltline mid-plane, the significant region of interest to provide low temperature overpressure protection.
This issue was not previously considered in the Westinghouse Owners Group (WOG) methodology (Reference
- 6) and was not specifically addressed in prior LTOP analyses although some unquantified margin was available.
The revised analysis specifically accounts for this pressure drop assuming the full flow of 3 RCPs and 2
Residual Heat Removal (RHR) pumps, and as before, the static head correction between the pressure sensor and the reactor vessel mid-plane location.
Also, the pressure sensor instrument uncertainties were increased due to newly obtained vendor data, test equipment assumptions and environmental effects.
As a result of the combined effects described
- above, the temperature ranges for the selected heatup and cooldown rates were extended downwards to lower temperatures (i.e., less restrictive) and are supported by analysis results.
As noted previously, the heatup and cooldown rates were not revised.
- However, the 10'F/hr.
cooldown rate has been extended for use below 115'F when LTOPS protection is enabled, provided less than 3 RCPs are in operation.
This benefits plant operation where control of temperatures is difficult (i.e., starting and stopping of RHR pumps and control of flowrates etc.).
This additional benefit is justified by the reduction in dynamic pressure drop across the reactor vessel due to reduction in flow rate associated with running less than 3 RCPs.-
As a consequence of a reduced dynamic pressure
- drop, the peak pressure of a mass input event (the most limiting event at low temperatures) is also reduced below the allowable pressure limit.
/
I II
ENCLOSURE TO SERIAL:
NLS-93-059 Instrument Uncertainties:
Instrument uncertainties for pressure and temperature were specifically accounted for in the LTOP analysis.
The proposed changes to TS Figure 3.4-4 footnote replace the instrument uncertainty values with a statement to the effect that the LTOP instrument uncertainties will be considered and the specific values to be used will be documented in Plant Procedure PLP-106, "Tech.
Spec.
Equipment List Program and Core Operating Limits Report".
PLP-106 identifies and controls the various equipment lists required by the SHNPP TS.
PLP-106 was generated concurrently with issuance of the SHNPP Full Power Operating License to fulfillthe requirements of various SHNPP TS.
Historically, those items comprising the TS Equipment List Program were maintained directly in the TS.
By maintaining these items independent of the TS, but still fulfillingthe regulatory requirements, they are maintained more current'nd accurate with respect to the actual plant configuration without the delay typically encountered with Operating License/Technical Specification amendments.
Requirements to maintain and use the information listed in PLP-106 is retained in the TS so that there is no adverse effect on the TS Limiting Conditions of Operation or the Surveillance Requirements.
The provisions of 10 CFR 50.59 provide adequate opportunity for review of changes to the information contained in the information listed in PLP-106 by the staff.
With regards the Instrumentation uncertainty currently accounted for in the LTOP setpoints in Figure 3.4-4, the specific values for uncertainty would be listed in PLP-106 and the Figure annotated to reflect documentation of the specific instrumentation uncertainties within PLP-106.
Regardless of where the magnitude of'he uncertainties is described, the plant operating procedures implement the heatup and cooldown rates including margin for instrument uncertainties.
For the same reason, the instrument uncertainties for pressure and temperature have been removed from TS Figures 3.4-2 and 3.4-3 regarding the heatup and cooldown curve pressure-temperature limits. It is proposed to control the pressure and temperature uncertainties for normal RCS pressure and temperature indication (and RHR heat exchanger outlet temperature) in Plant Procedure PLP-106.
Therefore, changes in instrument uncertainties due to, for example, sensor replacement will not require TS changes but will require plant procedure and design document changes.
A Conclusions The above demonstrates that utilizing the existing LTOP setpoints and design basis transients, the peak pressure developed as a result of overshoot beyond the setpoint will not exceed the revised Appendix G pressure-temperature limits, nor the reactor vessel closure flange material pressure-temperature limits, when used in conjunction with the proposed heatup and cooldown rates and applicable temperature ranges.
This provides assurance that the reactor
ENCLOSURE TO SERIAL:
NLS-93-059 vessel is protected from brittle fracture up to 11 EFPY.
ENCLOSURE TO SERIAL:
NLS-93-059 REFERENCES CP&L Letter to NRC, NLS-89-257, September 8,
1989, "Reactor Vessel Coupon Analysis Summary Technical Report."
(2)
CP&L Letter to NRC, NLS-92-097, April 2, 1992, 'Reactor Vessel Material Surveillance Report."
(3)
Regulatory Guide 1.99, Revision 2, May 1988, 'Radiation Embrittlement of Reactor Vessel Materials."
(4)
BAW-10046A, Revision 2, June 1986, Topical Report:
"Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G."
(5)
Welding Research Council (WRC) Bulletin 175, August 1972, 'VRC Recommendations on Toughness Requirements for Ferritic Materials'
(6)
Westinghouse Owners Group Report, July 1977, and its Supplement, September
- 1977,
" Pressure Mitigation Systems Transient Analysis Results."
(7)
Safety Evaluation Report, NUREG-1038, November 1983.
(8)
Safety Evaluation Report, NUREG-1038, Supplement 4, October 1986 (9)
NRC Letter to CP&L, May 31,
- 1990,
" Issuance of Amendment 19 to Facility Operating License No. NPF-63, SHNPP Unit 1, Regarding Pressure Temperature Limits Related to GL 88-11."
El-12
ENCLOSURE TO SERIAL:
NLS-93-059 ENCLOSURE 2 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS 10 CFR 50.92 EVALUATION The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists.
A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
Carolina Power
& Light Company has reviewed this proposed license amendment request and determined that its adoption would not involve a significant hazards determination.
The bases for this determination are as follows:
Pro osed Chan e
This Technical Specification change revises Specification 3.4.9 by replacing the current five EFPY heatup and cooldown limitations with revised limitations based on the predicted reactor vessel neutron exposure at 11 EFPY of operation.
The revisions affect the RCS pressure-temperature limitations and the effective ranges of the RCS heatup and cooldown rates.
In addition, the proposed amendment includes other related changes.
Basis This change does not involve a significant hazards consideration for the following reasons:
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Technical Specifications 3.4.9.1 and 3.4.9.2 "REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS" provide RCS pressure-temperature limits to protect the reactor pressure vessel from brittle fracture by clearly separating the region of normal operations from the region where the vessel is subject to brittle fracture.
The heatup and cooldown rates of Specifications 3.4.9.1 and 3.4.9.2, and LTOP setpoints in Specification 3.4.9.4 are designed to ensure that the 10 CFR 50 Appendix G
pressure-temperature limits for the RCS are not exceeded during any condition of normal operation including anticipated operational occurrences and system hydrostatic tests.
General Design Criterion 31 of Appendix A to 10 CFR 50 requires that the
ENCLOSURE TO SERIAL:
NLS-93-059 reactor coolant boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance,
- testing, and postulated accident conditions (l)the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
Title 10 of the Code of Federal Regulations Part 50 Appendix G, "Fracture Toughness Requirements,"
requires the effects of changes in the fracture toughness of reactor vessel materials caused by neutron radiation throughout the service life of nuclear reactors to be considered in the pressure-temperature limits'he change is used in conjunction with the material initial reference temperature (RT~z) to establish the limiting pressure-temperature curves.
Regulatory Guide 1.99 contains procedures for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.
Using the RG 1.99 Revision 2 and Appendix G to 10 CFR 50, new Adjusted Nil Ductility Reference Temperatures (ART~<) and limiting pressure-temperature curves were prepared for the projected reactor vessel exposure at ll Effective Full Power Years (EFPY) of operation.
These new curves in conjunction with the associated changes in the heatup and cooldown ranges and the existing Low Temperature Overpressure Protection System setpoints provide the required assurance that the reactor pressure vessel is protected from brittle fracture up to 11 EFPY of operation.
No changes to the design of the facility has been made.
No new equipment has been added or removed and no operational setpoints have been altered.
The revised analysis and resultant adjustment of the operating limitations provide assurance that the Reactor Coolant System is protected from brittle fracture.
Therefore, the proposed amendments to the pressure-temperature limitations, the heatup and cooldown ranges, and the recalculated limiting material ART~< do not involve a significant increase in the probability or consequences of an accident previously evaluated; collectively they maintain the required buffer necessary to protect the reactor vessel from brittle fracture given a limiting mass or temperature input to the RCS for up to 11 EFPY of operation.
2.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
No new equipment has been added or removed and no operational setpoints have been altered.
The revised analysis and resultant adjustment of the operating limitations provides assurance that the Reactor Coolant System is protected from brittle fracture.
No new accident or malfunction mechanism is introduced by this amendment.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
E2-2
ENCLOSURE TO SERIAL:
NLS-93-059 The proposed amendment does not involve a significant reduction in the margin of safety.
The heatup and cooldown rates of Specifications 3.4.9.1 and 3.4.9.2, and LTOP setpoints in Specification 3.4.9.4 are designed to ensure that the 10 CFR 50 Appendix G pressure-temperature limits for the RCS are not exceeded during any condition of normal operation including anticipated operational occurrences and system hydrostatic tests.
New Nil Ductility Reference Temperatures and limiting pressure-temperature curves were prepared for the projected reactor vessel exposure at 11 Effective Full Power Years of operation in accordance with 10 CFR 50 Appendix G and the methodology provided in Regulatory Guide 1.99, Revision 2.
The revised heatup and cooldown ranges, in conjunction with the current rates and LTOP setpoints ensure that the Appendix G pressure-temperature curves are not challenged given a limiting mass or heat input to the RCS during normal operations, anticipated occurrences and system hydrostatic testing.
Since restrictions remain in place to ensure the Appendix G operating limits of the reactor vessel are not challenged, the margin of safety defined in the Technical Specification BASES is not significantly reduced by this change.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
E2-3
ENCLOSURE TO SERIAL:
NLS-93-059 ENCLOSURE 3
SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS ENVIRONMENTAL CONSIDERATIONS 10 CFR 51.22(c)(9) provides criterion for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.
A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not:
(1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (3) result in an increase in individual or cumulative occupational radiation exposure.
Carolina Power
& Light Company has reviewed this request and determined that the proposed amendment meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment.
The basis for this determination follows:
Pro osed Chan e
This Technical Specification change revises Specification 3.4.9 by replacing the current five EFPY heatup and cooldown limitations with revised limitations based on the predicted reactor vessel neutron exposure at ll EFPY of operation.
The revisions affect the RCS pressure-temperature limitations and the effective ranges of the RCS heatup and cooldown rates.
In addition, the proposed amendment includes other related changes.
Basis The change meets the eligibilitycriteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:
As demonstrated in Enclosure 2, the proposed amendment does not involve a significant hazards consideration.
The proposed amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
The changes proposed are to administrative controls and limits unrelated to effluents generated or released from the facility.
As such, the change can not affect the types or amounts of any effluents that may be released offsite.
Page E3-1
ENCLOSURE TO SERIAL:
NLS-93-059 The proposed amendment does not result in an increase in individual or cumulative occupational radiation exposure.
The changes proposed are to administrative controls and limi.ts unrelated to personnel radiation exposure.
Therefore, the amendment has no affect on either individual or cumulative occupational radiation exposure.
Page E3-2
ENCLOSURE 4 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS PAGE CHANGE INSTRUCTIONS Removed Pa e
3/4 4-35 3/4 4-36 3/4 4-38 3/4 4-41 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-11 B 3/4 4-12 B 3/4 4-14 B 3/4 4-15 Inserted Pa e
viii 3/4 4-35 3/4 4-36 3/4 4-38 3/4 4-41 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-11 B 3/4 4-12 B 3/4 4-14 B 3/4 4-15