ML17262A377
| ML17262A377 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/15/1991 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17262A376 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 NUDOCS 9102250126 | |
| Download: ML17262A377 (19) | |
Text
ATTACHMENT A Revise the Technical Specification pages as follows:
Remove 3.1-10 3.1-11 3
~ 1 1 2 3 ~ 1 1 3 3.1-14 3.1-15 3.1-16
-3.1-16a J
F 1 17 3.1-18 3.3-4a 3+3 13 3.3-14 3.15-1 3.15-2 Insert 3.1-10 3.1-10a 301-11 3 ~ 1 1 2 301-13 3.1-14 3.3-4a 3.3-4b 3 ~ 3 1 3 3.3-14 3.3-14a 3.15-1 3.15-2 c
9i022 <Oi26 9i02i5 PDR ADOCK "05000244 P
~ 1 ~ 2
~
~
Heatu and Cooldown Limit Curves for Normal 0 eration 3.1.2.1 The reactor coolant temperature red pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.1-1 and 3.1-2 for the first 21.0 effective full power years.
a.
Allowable'ombinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
The heatup and cooldown rates shall not exceed 60'F/hr and 100'/hr, respectively.
Limit lines for cooldown rates between those presented may be obtained by interpolation.
b.
Figures 3.1-1 and 3.1-2 define limits to. assure
/
prevention of non-ductile failure only.
The limit lines shown in Figures 3.1-1 and 3.1-2 shall be recalculated periodically using methods discussed in the Basis Section.
- c. If the limits on Figures 3.1-1 and.
3.1-2 are
- exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; and either 1) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System and determine that the Reactor Coolant System remains acceptable for continued operation, or 3.1-10 Amendment. No.
15
2) within 6
hours be in at least HOT
- SHUTDOWN, and within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> reduce RCS temperature and pressure to less than 200'F and 500
- psig, xespectively.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator vessel is below 70'F.
The pressurizer heatup and. cooldown rates shall not exceed 100'F/hx and 200'F/hr, respectively.
The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320'F.
3.1-10a Amendment No.
15
Basl.s Fracture Tou hness Pro erties The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the Summer 1965 Section III of the ASME Boiler and Pressure Vessel
- Code, Reference (1),
- and, ASTM E185, Reference (2),
and. in accordance with additional reactor vessel requirements.
These properties are then evaluated in accordance with Appendix 6 of the 1972 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code, Reference (3) and the calculation methods described, in Reference (4).
The results are reported in Reference (5) for (
Capsule T.
The heatup and. cooldown curves are based on nominal pressure-temperature indications.
Sufficient conservatism exists in the algorithm from which the curves were derived to account for instrument uncertainties.
TEXT DELETED 3.1-11 Amendment No.
15 Proposed
The temperature requirements for the steam generator corresponds with the measured NDT for the shell of the steam generator.
A temperature difference of 320'F between the pressurizer and reactor coolant system maintains stresses within the pressurizer spray nozzle below design limits.
ASME Boiler and Pressure Vessel Code Section III (Summer 1965)
ASTM E185 Surveillance Tests on Structural Materials in Nuclear Reactors ASME Boiler and Pressure Vessel
- Code,Section III, Summer 1972 Addenda (note Code Class 1514)
Regulatory Guide 1.99, Rev.
2, May 1988, "Radiation, Embrittlement of Reactor Vessel Materials."
- Report, "Rochester Gas and Electric Reactor Vessel Life Attainment Plan",
dated March 1990.
3.1-12 Amendment No.
15 Proposed
RCS PRESSURE (PSIG) 2500 2250 LIMITING MATERIAL CIRCUMFERENTIAL VELD LIMITING 1/4T RTndt=210'F LIMITING 3/4T RTndt=178'F HYDRO LIMIT 2000 1750 1500 UNACCEPTABLE OPERATION 1250 1000 60 oF CRITICALITY LIMIT 750 500 ACCEPTABLE OPERATION 329'F 250 50 100 150 200 250 300 350 400 450 500 RCS TEMPERATURE (DEG F)
Figure 3.1-1:
Ginna Reactor Vessel Heatup Limitations Applicable for the first 21 EFPY using Reg Guide 1.99, Rev.
2
- 3. 1-13 Amendment No.
15
I' 1
4 t,
1
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~ l(
v I
1t P
l
~ 4>"
h
~
t
RCS PRESSURE (PSIG) 2500 2250 LIMITING MATERIAL CIRCUMFERENTIAL WELD LIMITING 1/4T RTndt=210'F LIMITING 3/4T RTndt=178'F
, 2000 1750 1500 UNACCEPTABLE OPERATION 1250 1000 750 500 250 COOLDOWN RATES O'P/HR 20'F/HR 40'P/HR 60'F/HR 100'F/HR ACCEPTABLE OPERATION 50 100 150 200 250 300 350 400 450 500 RCS TEMPERATURE (DEG F)
Figure 3.1-2:
Ginna Reactor Vessel Cooldown Limitations Applicable for the first 21 EFPY using Reg Guide 1.99, Rev.
2.
- 3. 1-14 Amen~ment No.
15
v 3 ~ 3
~ 1 e7 3.3.1.7.1 3.3.1.8 3.3.1.8.1 Except during diesel generator load and safeguard sequence testing or when the vessel head is removed, or the steam generator primary system manway is open, no more than one safety injection pump shall be operable whenever the overpressurization protection is provided by a RCS vent of > 1.1 sq. in. (3.15.1.b).
Whenever only one safety injection pump may be operable by 3.3.1.7, at least two of the three safety injection pumps shall be verified inoperable, as defined in the Basis for this section, a minimum of once per twelve hours.
Except during diesel generator load and safeguard sequence testing or when the vessel head is removed, or the steam generator primary system manway is open, all three safety injection pumps shall be inoperable and safety injection discharge paths to the RCS isolated whenever overpressure protection is provided by the pressurizer PORVs (3.15.1.a).
Whenever safety injection pumps are required to be inoperable by 3.3.1.8, the safety injection pumps shall be verified inoperable, as defined in the Basis of this section, a minimum of once per twelve hours.
Similarly safety injection discharge paths to the RCS shall be verified to be isolated a minimum of once per twelve hours.
3.3-4a Proposed.
3.3.1.8.2 The requirements of 3.3.1.8 may be modified to allow operation of one SI pump provided the associated paths to the RCS are isolated by A.C. power being removed to the discharge MOVs in the closed position, or the manual isolation valves closed.
Isolation of the discharge paths shall be verified at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.3-4b Proposed
The facility has four service water pumps.
Only one is needed during the injection phase, and two are required during the recirculation phase of a
postulated loss-of-coolant accident.'
The control room emergency air treatment system is designed to filter the control room atmosphere during periods when the control room is isolated and to maintain radiation levels in the control room at acceptable levels following the Design Basis Accident.'
Reactor operation may continue for a
limited time while repairs are being made to the air treatment system since it is unlikely that the system would be needed.
Technical Specification 3.3.5 applies only to the equipment necessary to filter the control room atmosphere.
Equipment necessary to initiate isolation of the control room is covered by another specfication.
The limits for the accumulator pressure and volume assure the required amount of water injection during an accident, and are based on values used for the accident analyses.
The indicated level of 50:
corresponds to 1108 cubic feet of water in the accumulator and the indicated level of 82% corresponds to 1134 cubic feet.
The limitation of no more than one safety injection pump to be operable when overpressure protection is being provided by a RCS vent of ) 1.1 sq. in. insures 303-13
that the mass addition from the inadvertant operation of safety injection will not result in RHR system pressure exceeding design limits.
The limitation on no safety injection pumps operable and the discharge lines isolated when overpressure protection is provided by the pressurizer PORV's removes mass injection from inadvertant safety injection as an event for which this configuration of overpressure protection must be designed to protect.
Inoperability of a
safety injection pump may be verified from the main control board with the pump control switch in pull stop, or the pump breaker in the test or racked, out position such that the pump could not start from an inadvertant safety injection signal.
Isolation of a
safety injection pump discharge path to the RCS may be verified from the main control board by the discharge NOV switch position indicating closed, or the discharge valve closed with A.C. power removed, or a manual discharge path isolation valve closed such that operation of the associated safety injection pump would not result in mass injection to the RCS.
The limitation on boric acid. storage tank volume is based on the assumption that 2000 gallons of 12% to 13% solution is delivered to the RCS during a large steam line break associated with the recirculation flow to the RWST and the sweep volume in the SI pump Amendment No. 24, 33 3.3-14
suction line and still deliver 2000 gallons to the RCS.
References I
(1) Deleted (2)
UFSAR Section 6.3.3.1 (3)
UFSAR Section 6.2.2.1 (4)
UFSAR Section 15.6.4.3 (5)
UFSAR Section 9.2.2.4 (6)
UFSAR Section 9.2.2.4 (7)Deleted.
(8)
UFSAR Section 9.2.1.2 (9)
UFSAR Section 6.2.1.1 (Containment Integrity) and UFSAR Section 6.4 (CR Emergency Air Treatment)
(10) Westinghouse
- Analysis, "Report for the BAST Concentration for R. E. Ginna", August 1985 submitted by RG&E letter from R.
W. Kober to H. R. Denton, dated October 16, 1985.
3.3-14a
3.15 Over ressure Protection S stem A licabilit Applies whenever the temperature of one or more of the RCS cold legs is
< 330'F, or the Residual Heat Removal System is in operation.
Ob'ective To prevent overpressurization of the reactor coolant system and the residual heat removal syst: em.
S ecification 3.15.1 Except during secondary side hydrostatic tests in which RCS pressure is to be raised above the PORV setpoint, at least one of the following over-pressure protection systems shall be operable:
a.
Two
. pressurizer power operated relief valves (PORVs) with a liftsetting of < 424 psig, or b.
A reactor coolant system vent of
> 1.1 square inches.
3.15.1.1 With one PORV inoperable, either restore the inoperable PORV to operable status within. 7 days or depressurize and vent the RCS through a 1.1 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to operable status.
3.15.1.2 With both PORVs inoperable, depressurize and vent the RCS through a 1.1 square inch vent(s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented. condition until both PORVs have been restored to operable status.
3.15.1.3 Use of the overpressure protection system to mitigate an RCS or RHRS pressure transient shall be reported in accordance with 6.9.2.
Basis An RCS vent opening of greater than 1.1 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are
< 330'F'.
This relief capacity will 3.15-1 Proposed
ensure that no overpressurization of the RHR system could occur.
The vent opening protects the RCS and RHRS from overpressurization when the transient is limited to either 1) the start of an idle RCP with the secondary water temperature of the steam generator
< 50'F above the RCS cold leg temperature or
- 2) the start of a safety injection pump and its injection into a water solid RCS'he operability of two pressurizer PORVs ensures that the RCS will be protected from pressure transients which could exceed the. limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are
< 330'F'~'.
This relief capacity willalso ensure that no overpressurization of the RHR system could occur.
Either PORV has adequate relieving capability to protect the RCS and RHRS from overpressurization when the transient is limited to either 1) the start of an idle RCP with the secondary water temperature of the steam generator
< 50'F above the RCS cold leg temperature or 2) charging/letdown mismatch with three charging pumps in operation'eferences:
L. D. White, Jr., letter to A. Schwencer, NRC, dated July 29, 1977.
(2)
SER for SEP Topics V-10.B, V-11.B, VII-3, "Safe Shutdown," dated September 29, 1981.
(3)
- Report, "R.
E.
Ginna Low Temperature Overpressure Protection System (LTOPS) Setpoint Phase II Evaluation Final Report,"
dated.
February 1991 submitted by letter to Allen R. Johnson, NRC, dated February 15, 1991.
3.15-2 Amendment. No. 1, 11 Proposed
Attachment B
G neric Letter 88-11 (Ref.
- 1) transmitted a copy of Revision 2 to Regulatory Guide 1.99.
GL 88-11 required each Licensee to re-evaluate the effect of neutron radiation on reactor vessel material using the methods in Revision 2. If the re-evaluation resulted in more restrictive heatup/cooldown curves the new curves should be used to revise Technical Specifications.
The re-evaluation resulted in new curves for R.
E.
Ginna.
The new curves are presented on Figures 3.1-1 and 3.1-2 in Attachment A.
The curves on Figure 3,.1-1 and 3.1-2 were generated using the methodology presented in Revision 2 to Reg. Guide 1.99.
The curves are applicable for 21.0 EFPY.
Currently Ginna has accumulated approximately 15.4 EFPY.
The curves are based on nominal values, i.e. the pressure and temperature instrumentation uncertainties are not included.
This is consistent with standard practice.
The requisite conservatism has been sufficiently addressed by the algorithm from which Appendix G limits are derived.
The new heatup/cooldown curves required a re-evaluation of the Low Temperature Overpressurization Protection System (LTOPS) setpoint.
The evaluation is presented in Attachment C.
An LTOPS setpoint of 424 psig or less will provide overpressure protection for both the RHR System and the 21 EFPY isothermal Appendix G limits.
The setpoint is based on an evaluation of mass input events and heat input events while in the water solid condition.
The most limiting mass injection case is the charging/letdown mismatch
'with three positive displacement charging pumps.
Technical Specification limits on SI pump operability and discharge valve position eliminate the mass injection case due to a high head SI pump.
With three SI pumps inoperable an inadvertent SI signal will not cause a pump start.
With the SI system discharge paths isolated no single inadvertent operator action will result in SI.
Therefore, SI will not result from inadvertent operation.
The most limiting heat injection transient results from restarting a
RCP with the steam generator secondary side water and primary side tube water 50'otter then the rest of the RCS.
The Technical Specification changes associated with this Amendment are:
1)
Revised heatup/cooldown curves based on Rev.
2 to Reg.
Guide 1.99 methodology.
2)
~
~
Deletion of Basis material which is more appropriatly documented in educational material on how heatup/cooldown curves are generated.
(See Table 1 attached).
3)
- Revised, LTOPS setpoint.
4)
Additional Technical Specification restrictions necessary to eliminate SI injection as a possible mass injection case when the PORVs are used. to provide overpressure protection.
These changes to the Technical Specifications, in accordance. with 10 CFR 50.91, have been evaluated to determine if the operation of the facility in accordance with the proposed Amendment would:
1)
Involve a
significant increase in the probability or consequences of an accident previously evaluated; or 2)
Create the possibility of-a new or different kind of accident from any accident previously evaluated; or 3)
Involve a significant reduction in the margin of safety.
Re-calculation of heatup/cooldown limits using approved NRC methodology does not increase the probability or consequences of an accident.
New limits do not create a different kind of accident.
New limits generated from approved methodology do not reduce the margin of safety.
Deletion of education material that is more appropriately documented outside of Technical Specification does not effect the probability consequences, or margin of safety of an accident.
Neither does it cause a different type accident.
Reducing the LTOPS setpoint protects the Appendix G limits
- and, prevents RHR overpressurization.
A setpoint, reduction reduces the consequences of an overpressurization event because mitigation is initiated. earlier.
This results in a smaller challenge to the RHR system and reduces the probability of failure.
Reducing the LTOPS setpoint does not cause a different kind of accident.
Since existing acceptance. criteria are still met there is no reduction in a margin of safety.
Additional restrictions ensure no single inadvertent operator action willresult in mass injection due to SI.
The LTOPS setpoint based.
on possible mass injection and heat injection cases is adjusted.
to meet existing acceptance criteria.
Therefore, the consequences or probability of an accident are not increased, a new accident is not created or is a margin of safety reduced.
Reference (1)
NRC Generic Letter 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and, Its Impact on Plant Operations,"
dated July 12, 1988.
Table 1
Detail Description of Changes Section Chan e
Reason for Chan e
Page 3.1-3.0 added item c for action if heatup/cooldown limits are exceeded additional guidance Page 3.1-11 Paragraph 1
Page 3.1-11 added "The results are reported in Reference (5) for Capsule T.
Basis for nominal
.values Document source of heatup/cooldown curves additional, information Page 3.1-11 Paragraphs 2 and 3
Page 3.1-12 through 3.1-16 Page 3.1-16a Paragraph 1 and 2
Page 3.1-16a References Deleted Deleted Deleted Reference 4 and 5
changed (4) Reg.
Guide 1.99, Rev.
2, May 1988 "Radiation Embrittlement of Reactor Vessel Material" Material not approp-riate for Tech.
Spec.
Material not approp-riate for Tech.
Spec.
Material not approp-riate for Tech.
Spec.
Document source of heatup/cooldown curves and make Ref.
consistent with text (5) Westinghouse
- Report, "RGGE Reactor Vessel Life Attain-ment Plan", dated March 1990 Reference 4 deleted Page 3.1-17 Page 3.1-17 changed to 3.1-13 New heatup curve Page 3.1-18 Figure 3.1-1 changed to new heatup curve Page 3.1-18 changed to 3.1-14 New cooldown curve Figure 3'-2 changed to new cooldown curve
Section Chan e
Reason for Chan e
Page 3.3-4a Section 3.3.1.7 I
Page 3.3-4a system is required to be operable."
changed to ". ~. is provided by a RCS vent of 2 1.1 sq. in.
(3.15.1.b)
Added Section 3.3.,1.8 Separate requirement associated with vent from requirements associated with PORVs Equipment restrictions associated with overpressure protection via the PORVs is more restrictive than with a vent.
Page 3.3-4a New Section 3.3.1.8.1 and 3.3.1.8.2 added Allow operation of SI pump and allow testing/maintenance of valves 878A, 878B, 878C and 878D.
Page 3.3-13 added Basis for 1.1 sq.
in. vent and PORVs additional information Page 3.2,5-1 Section 3.15.la Page 3.15-1 and 3.15-2 435 psig changed to 424 psig Changed basis to reflect assumptions used in analysis associated with protection by the vent or by the PORVs New LTOPS setpoint Revise basis