ML20236X115
| ML20236X115 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/01/1987 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp |
| Shared Package | |
| ML20236X116 | List: |
| References | |
| DPR-50-A-135 NUDOCS 8712080430 | |
| Download: ML20236X115 (18) | |
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. UNITED STATES
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- NUCLEAR REGULATORY COMMISSION.
.4 f E WASHINGTON, D C. 20555 g
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/e METROPOLITAN EDISON COMPANY JERSEY CENTRAL POVER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC C0tiPANY GPU NUCLEAR CORPORATION
' DOCKET NO. 50-289
.THREE MILE ISLAND NUCLEAR STATION, UNIT NO.-1
. AMENDMENT TO FACILITY OPERATING LICENSE-Amendment. No.135.
License No. DPR-50 l
~.1..
The Nuclear Regulatory Comission (the Comission)' has found that':
A.
The application for amendment by GPU Nuclear Cor) oration, et.al.
(the licensee) dated March 5,.1987, complies wit 1 the tanda'rds and req)uirements.of the Atomic Energy Act of 1954, as s
amended (the Act, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will oserate in conformity with the application, the provisions of t1e Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this. amendment can be' conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations;
'D.
The issu6nce of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.-
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have_been satisfied.
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Accordingly, the~. license is amended by changes to the Technical l
Specifications as indicated in.the attachment to this. license
. amendment, and. paragraph 2.c.(2) of Facility Operating License
'No. DPR-50 is hereby amended to reac' as follows:
l Technical Specifications i
The Technical Specifications contained in Appendix A, as revised through. Amendment 110.135, are hereby incorporated.in the license. GPU Nuclear Corporation shall operate.the' facility in accordance with the Technical Specifications.
'3.
.This: license amendment is effective as of its date of issuance.
FORTHENUCLEARREGhLATORYCOMMISSION sl. l (JohpF.Stolz, Direct'or Project Directorate I-4 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
December ~ 1.1987
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ATTACHMENT TO LI.C.E.N.S.E..N.I.E.N.D_ MEN.T..N.O 135 l
FACILITY' OPERATING' LICENSE NO. DPR '
DOCKET NO. 50.289 Replace thi'following'pages of.:t.he Appendix:"A" Technical. Specifications
- with.the: attached'pages. The. revised pages are idertified bp Arrendiner.t :
q nurber and'contain vertical-lines -indicating the area'cf change..
- Rer,cVe '
Insert 2 '2-4 2-7
~2-7 2-8:
2-8 2-g 2-9 Figure 2.3-1
-Figure 2.3-1 3-27a 3-27a 3-28a 3-28a lc 3-30 3-30.
l 4-7.
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2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE Applicability i'
Applies to the limit on reactor coolant system pressure.
Ob_iective To maintain the integrity of the reactor coolant-system and to prevent the release of significant amounts of fission product activity.
Specification.
2.2.1-
.The reactor' coolant system pressure shall not enceed 2750 psig when
.there are fuel assemblies in the reactor vessel.
Bases
'The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching.the atmosphere.
In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the'eeactor coolant system. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III, is 110% of design pressure (2). The maximum' transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110% of' design pressure.
Thus, the safety limit of-2750 psig (110% of the 2500 psig design pressure) has been established (2). The.
maximum settings for the reactor high pressure trip (2355 psig) and the pressurizer code safety valves.(2500 psig)(3) have been established in accordance with ASME Boiler and Pressure Vessel Code Section III, Article 9 Winter, 1968 to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test was conducted at 3125 psig (125% of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by the presence of a pressurizer electromatic relief valve (4).
References (1) FSAR, Section 4 (2) FSAR, Section 4.3.10.1 (3) FSAR, Section 4.2.4 (4) FSAR, Table 4-1 3
I i
1 2-4 Amendment No. 12, 28, 39, 45, M, 135 l
c.
Re:ctot-coolcnt systcm pressura
{
During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint.
The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure ensures that'the system pressure is maintained below the safety limit (2750 psig) for any design transient (6). Due to calibration and instrument errors, the safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.
As part of the post ~1NI-2 accident modifications, the high pressure trip setpoint was lowered from 2390 psig to 2300 psig.
(The FSAR Accident Analysis Section still uses the 2390 psig high pressure trip setpoint.)
The lowering of the high pressure trip setpoint and raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORV while maintaining ASME Code Safety Valve capability.
A B&W analysis completed in September of 1985 concluded that the high reactor coolant system pressure trip setpoint could be raised to 2355 psig with negligible impact on the frequency of opening of the PORV during anticipated overpressurization transients (8).
The high pressure trip setpoint was subsequently raised to 2355 psig.
The potential safety benefit of this action-is a reduction in the frequency of reactor trips.
The low pressure (1800 psig) and variable low pressure (11.75 Tour-5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4).
The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig.
Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trip setpoints.
d.
Coolant outlet temocrature The high reactor coolant outlet temperature trip setting limit (618.8F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperature in the operating range.
The calibrated range of the temperature channels of the RPS is 520' to 620*F.
The trip setpoint of the channel is 618.8F. Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is 1.2*F.This accuracy was arrived at by summing the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefore, it is assured that a trip will occur at a value no higher than 620*F even under worst case conditions. The safety analysis used a high temperature trip setpoint of 620*F.
The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc.
This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 107. above the calibrated range.
2-7 Amendment No. 77, 2E, 39, 45, 7E, JM,135
Sine 9.it has bssn estsblish2d.that the chtnn21'will trip ct a'valu2 of RC outist tcmpsrotura no highsr th:n 620*F evsn in tha worst ecse, and since the channel is fully operational approximately 10% above'the calibrated E
range.and exhibits no hysteresis or foldover characteristics, it is concluded that the instrument design is' acceptable.
e.
Reactor building pressure The high reactor building pressure-trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
f.
Shutdown bypass In order to provide for control rod drive tests, zero power physics testings, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1.
Two conditions are imposed when the bypass is used:
By administrative control the nuclear overpower trip setpoint must be reduced to value $i 5.0 percent of rated power during reactor shutdown.
2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed..
This high pressure trip' set point is lower than the. normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The overpower trip setpoint of j[ 5.0 percent prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
References (1) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4).FSAR, Section 14.1.2.8 (5) FSAR, Section 14.1.2.6 (6) Technical Specification Change Request No. 31. January 16, 1976, and Technical Specification Change Request No. 84, June 23, 1978.
(7) "ECCS Analysis of B&W's 177-FA Lowered Loop NSS," BAW-10103, Rev. 2, Babcock and Wilcox, April 1976.
(8) " Justification for Raising Setpoint for Reactor Trip on High Pressure",
BAW-1890, Rev. O, Babcock and Wilcox, September 1985.
2-8 i
Amendment No. 45, J4I,135
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- Amendment No. 13,17, 28, 39, '45, 7E, Jd,135
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V 3.5.1.7.1 Power may be restored through' the breaker with the failed trip. feature for up to two hours for surveillance testing 7
per T.S? 4.1.1.
P 3.5.1.8
' During' STARTUP, HOT STANDBY or POWER. 0PERATION, in the:
event that one of the two regulating control rod power SCR-u electronic tripsois inoperable,, within one. hour:
a.: P1 ace:the inoperable.SCR electronic trip in the' tripped condition or
-b
-Remove:the power supplied to the associated SCRs.
- Specification 3.0.1 ' applies.
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. 3.5.1.8.1 Power.may be restored through the SCRs with the failed -
. electronic trip for up to two hours for surveillance testing per.T.S. 4.1.1.
3.5.1.9 The reactor shall not be in the Startup mode or in a critical : state unless both HSPS actuation logic trains
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associated with the Functional units listed in Table 3.5-1 are operable except as provided in Table 3.5-1,D.
3.5.1.9.1, With; one HSP5 actuation' logic train inoperable, restore the' train' to OPERABLE or' place. the inoperable device in an actuated. state ~within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With both HSPS actuation logic' trains i
inoperable, restore.one train to;0PERABLE within 1: hour.or be in HOT SHUTDOWN within the next-6 hours.
Bases Every reasonable effort will be made' to maintain all safety.
instrumentation in operation. A startup is not' permitted unless-three power range neutron. instrument chsnnels and two channels each of the following.are operable:
four reactor. coolant temperature instrument channels, four reactor coolant flow instrument channels,
'.four reactor coolant pressure instrument channels, four pressure-
. temperature instrument channels four. flux-imbalance; flow. instrument channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The reactor trip, on loss of.feedwater may be bypassed below 7% reactor power.
the bypass is automatically removed when reactor power is raised above 7%. The reactor-trip, on turbine trip, may be bypassed below 45% reactor power. The satety feature actuation system must have l
two, analog channels functioning correctly prior to startup.
The anticipatory reactor trips on loss of feedwater pumps and turbine trip have been added to reduce the number of challenges-to 9 : *9 the~ safety valves and power operated relief valve but have not been
' crediteo in the safety analyses.
,0peration at rated power is permitted as long as the systems have
'at least the' redundancy requirements of Column "B" (Table 3.5-1).
LThis-is in agreement with redundancy and single failure criteria of 4
IEEE 279 as. described in FSAR Sect W 7.
- There are four reactor protection chwels.
Normal trip logic is
- two' out of four. ' Required trip log 1c for the power range instrumen-tation. channels is two out' of: three.
Minimum trip logic on other instrumentation channels is one-out of two.
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HSPS instrument operability specified meets the single failure criterion for the EFW system.
Four instrument channels are provided for automatic EFW initiation on OTSG low level and high reactor building pressure, and for automatic main feedwater isolation on low OTSG pressure. Normal trip logic is two out of four.
With one of the 4 channels in bypass, a second channel may be taken out of service (placed in the tripped position) and no single active failure will prevent actuation of the associated HSPS train actuation logic.
No single active failure.of either HSPS train will prevent the other HSPS train from operating. to supply EFW to both OTSGs.
REFERENCE (1) FSAA, Section 7.1 (2)
" Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip," BAW-1893, Rev. O, Babcock & Wilcox, October 198S.
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i ENCLOSURE 1 EG&G IDAHO REVIEW OF TECHNICAL SPECIFICATION CHANGE REQUEST NO 154 FOR THE GPU NUCLEAR CORPORATION PLANT THREE-MILE ISLAND UNIT 1 DOCKET NO 50-289 OPERATING LICENSE NUMBER DPR-50
1.0 INTRODUCTION
In a letter from H. D. Hukill, GPU Nuclear Corporation (GPU), to the U.S.
Nuclear Regulatory Comission (NRC) dated March 5,1987 (Reference 1), GPU proposed license amendments to Facility Operating License DPR-50 for Three Mile Island Unit 1 (TMI-1). The proposed amendment would:
(1) increase the setpoint for trip (i.e., shutdown) of the reactor on high pressure in the reactor coolant system from 2300 psig to 2355 psig; and (2) increase the arming threshold for anticipatory reactor trip (ART) on turbine trip from 20% of full pcwer to 45% of full power.
In respnnse to the NRC request, GPU provided additional supporting information by letter, H. D. Hukill to the NRC dated August 4, 1987 (Reference 2). This report presents the EG&G Idaho review of the proposed amendment and the supporting information.
2.0 BACKGROUND
The Babcock and Wilcox (B&W) Nuclear Steam Supply System (NSSS) was designed with the capability to adjust to minor plant upsets and certain anticipated events such as feedwater transients, rapid load changes and turbine trips without a reactor trip. The system was designed to initiate a plant runback, upon detection of an upset or equipment malfunction, to a power level consistent with the plant condition and to limit the rise in Reactor Coolant Syrtem (RCS) pressure to less than the reactor trip setpoint by opening the EG&G DOCUMENT 1
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E pressurizer Power Operated Relief Valve (PORV). Subsequent to the TMI-2 accident, the Nuclear Regulatory Commission (NRC), by IE Bulletin 79-05B IReference 3), required licensees for all B&W Pressurized Water Reactor (PWR1 facilities to make modifications to reduce the number of automatic actuation of the PORVs. The modifications proposed by B&W on behalf of the owners group and accepted by the NRC included (1) raising the PORV.setpoint from 2255 psig to 2450 psig, (2) decreasing the reactor trip on high RCS pressure from 2390 psig to 2300 psig and (3) providing an ART for turbine trips above 20% power.
In addition, the NRC required that B&W demonstrate that these modifications:
(1) limited the frequency of PORV openings to less than 5% of the total number of overpressure transients (NUREG 0737 Item II.K.3.7 Reference 4) and (2) limited the probability of a small-break Loss of Coolant Accident (LOCA) caused by a stuck-open PORY to less than.001 per reactor-year (NUREG 0737 Item II.K.3.2 Reference 4). B&W submitted a report (Reference 5) to demonstrate that the modifications did meet the requirements. The NRC issued a Safety Evaluation Report (Reference 6) which concluded that the requirements were met.
Although these modifications have met the objectives of reducing challenges to and opening of the PORV, they have increased the frequency of reactor rips.
Each reactor trip results in a challenge to plant safety systems and any reduction in reactor trip frequency will contribute to overall plant safety as well as plant availability. On behalf of the B&W Owners Group, B&W has submitted two reports:
(1) " Justification for Raising Setpoint for Reactor Trip on High Pressure," BAW-1890. September 1985 (Reference 7) and (2) " Basis for Raising Aming Threshold for Anticipatory Reactor Trip on Turbine Trip,"
BAW-1893, October 1985 (Reference 8). These reports present the B&W justification for raising the high pressure reactor trip setpoint from 2300 psig to 2355 psig and increasing the ART for turbine trips from 20% power to 45% power. These reports state that with these changes, the NRC requirements for limiting the frc.luency of PORY openings and limiting the probability of a s.nall-break LOCA due to a stuck open PORV will still be met.
Further, these reports state that with these changes, the number of reactor trips per reactor year will be reduced by approximately 10%.
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i In April 1986,_ the NRC staff completed its review of these two B&W reports. -In their Safety Evaluation Reports (SERs), the' staff:
(1) reviewed the basis for the proposed changes;-(2) reviewed B&W's method of analysis of the effect of the proposed high pressure trip setpoint on'PORY openings; (3) compared the results of Monte Carlo simulation for PORV openings with the NRC requirements contained in NUREG-0737 (Reference 4); and (4) reviewed the results of B&W's analysis of the arming threshold for ART. The NRC requirements include:
(1) the PORV will open in less than 5% of all anticipated overpressure transients (NUREG 0737. Item II.K.3.7 Reference 4); and (2) the probability of a small-break loss of coolant accident caused by a stuck-open PORV will be less than 0.001 per reactor-year (NUREG 0737. Item II.K.3.2 Reference 4).
In the SERs (References 9 and 10), the staff concluded on a generic basis that the l
proposed changes met the NRC requirements, and should benefit plants by potentially reducing the reactor trip frequency. Accordingly, the NRC concluded that the B&W reports could be referenced in licensing submittals by
.the B&W Owners Group members.
i 3.0 EVALUATION l
The licensee is a member of the B&W Owners Group and in their March 5, 1987 proposal (Reference 1), referenced the B&W reports BAW-1890 (Reference 7) and BAW-1893 (Reference 8) as justification for raising the high pressure reactor trip setpoint from 2350 psig to 2355 psig and increasing the ART for turbine trip to 45% power. The licensee states these reports are applicable to TMI-1 because:
(1) 1MI-1 is a B&W 177FA plant type for which the reports apply, (2) the mix of transients used to develop the historical data and the pressure overshoot distribution included data from TMI-1 for pre-TMI and similar B&W plants for post-TMI, (3) t.he actual measured high pressure reactor trip setpoint of 2294 psig (nomnally set at 2300 psig) and the maximum expected string error at the pressure bistable of 3.52 psi for TMI-1 are within the range of the plants used to develop the statistical data and (4) the available steam by-pass, including the first bank of main steam safety valves, is 76.9%
l-of full steam flow which is conservatively higher than the 43% used in the analysis. Based on the infonnation provided and a review of the B&W reports, 1
EG&G concluded that data and parameters used in the B&W reports are a
(
conservative bound for TMI-I and the use of these reports for TMI-I is EG&G DOCUMENT 3
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f i
1 acceptable.
l 1
The important conclusion of the B&W reports which justifies raising the high pressure trip and the ART for turbine trip is that even though both changes will result in a small increase in the probability of opening the PORV, the increase is insignificant compared to the total openings of the PORY from all events. The estimate made in the B&W reports (References 7 and 8) for PORV I
openings from high pressure events wit'n the reactor high. pressure. trip set at 2355 psig and the ART for turbine trip set at 45% is 1.96 x 10-5 PORV openings per reactor year. The estimate made in the B&W report (Reference 6) for PORY l
openings from all events is 8.06 x 10-2 PORV openings 'per reactor year with the major contributor to the PORV openings identified as actions by the operators in carrying out plant operation in accordance with the Ab ormal Transient.
Operating Guidelines-(ATOG). The 1.96 x 10-5 PORV openings per reactor year from high pressure events applies to TMI-1 as discussed in the preceeding paragraph. The 8.06 x 10-2 PORV openings per reactor year from all events applies to TMI-1 because the TMI-1 plant specific Abnormal Transient Procedures
'(ATPs) are based on the AT0G. Therefore, for TMI-1, the PORY openings from
-high' pressure events with the proposed setpoints are insignificant compared with the openings from all events.
1 GPU states (Reference 1) that the proposed changes are within the bounds of the current TMI-1 accident analysis of Chapter 14 of the Final Safety Analysis Report (FSAR). These analyses use the initial high pressure trip of 2390 psig for evaluating high pressure events, which is a conservative bound to the proposed 2355 psi. The licensee also states (Reference 2) that the FSAR analyses do not take credit for the ART on turbine trip and are therefore still valid and a conservative bound for the proposed ART cn turbine trip of 45%
power.
I The licensee in their letter of August 4, 1987 (Reference 2) confirmed that with the proposed setpoint changes more Main Steam Safety Valves (MSSVs) would i
l open during an overpressure transient or a turbine trip capared to those opening with.the current settings. However, they contend that with the proposed changes there would be fewer overpressure transients and turbine trips l
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-}
'that challenge the MSSVs and the net number _ of openings would be expected to be ireduced. -Also', the MSSVs-have a good record'of performance. Data from TMI-1
.and similar B&W plants show the MSSVs have lifted many times without failures I
to reseat. Based on the information provided by the licensee and the infomation from a similar plant, EG&G concurs that the anticipated openings of the'MSSVs are acceptable, i
3.0 CONCLUSION
S l
EG&G has reviewed.the proposed changes to TS 2.2 and Table 3.5-1 for TMI-1.
The' proposed changes would (1). increase the setpoint.for trip of the reactor on high' pressure in the reactor coolant system from 2300 psig to 2355 psig; and (2) increase the aming threshold for ART on turbine trip from 20% of full power to 45% of full power. As discussed in the preceding section, the EG&G finds the proposed changes meet the NRC positions established in their review of the B&W topical reports and therefore meet the applicable regulatory guidance and requirements and are,-therefore, acceptable.
l REFERENCES 1.
Letter from H. D. Hukill, GPU Corporation, to U.S. Nuclear Regulatory Commission. Letter number 5211-87-2053 dated March.5, 1987.
I 2.
Letter from H. D. Hukill, GPU Corporation, to U.S. Nuclear Regulatory Commission, Letter number 5211-87-2142 dated August 4, 1987.
j 3.
IE Bulletin 79-05B," Nuclear-Incident at Three Mile Island'- Supplement". April 2, 1979.
- 4. -
NUREG-0737, " Clarification of TMI Action Plan Requirements".
November 1980.
5.
Babcock & Wilcox Report 12-1122779 Rev. 1. " Report on PORY Opening Probability and Justification for Present Systems and Setpoints".
January 1981.
l' 6.
NRC Memorandum from F. H. Rowsome to G. C. Laines, " Safety Evaluation of the B&W Licensees' Responses to TMI Action II.K.3.2".
August 24, 1983.
7.
BAW-1890. " Justification for Raising Setpoint for Reactor Trip on High Pressure." September 1985.
v EG&G DOCUMENT 5
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_.J
8.:
BAW-1893, " Basis for Raising Arming Threshold.for Anticipatory Reactor Trip on Turbine Trip". 0ctober 1985.
9.
Letter from D. H. Crutchfield, NRC, to J. H. Taylor, B&W, Acceptance for' Referencing of Licensing Topical Report BAW-1890, " Justification for Raising Setpoint for Reactor Trip on High Pressure". April 27, 1986.
10.
Letter from D. M. Crutchfield, NRC,'to J. H. Taylor, B&W,: Acceptance for Referencing of Licensing Topical-Report BAW-1893 " Basis for Raising Threshold for Anticipatory Reactor Trip on Turbine Trip".
April 25, 1986.
W 4
ll' EG&G DOCUMENT 6
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