ML19319E230
| ML19319E230 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/19/1976 |
| From: | Goller K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19319E225 | List: |
| References | |
| NUDOCS 8003310737 | |
| Download: ML19319E230 (24) | |
Text
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NUCLEAR REGULATORY COMMISSION j
j wAsHWOTON, D. C. 2sses o
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SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO. 50-312 RANCHO SECO NUCLEAR CENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 4 License No. DPR-54 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applications for amendment by Sacramento Municipal Utility District (the licensee) dated July 8, 1975, as supplemented, May 23, 1975, and March 10, 1976, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with.the applications, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) 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 i
compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the conson defense and security or to the health and safety of the public; and E.
After weighing the environmental aspects involved, the issuance of this amendeent'is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable i
requirements have been satisfied.
g 3310
- 2. - Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license facendment.
13.
This license amendment is effcetive as of the date of its issuance.
TOR THE NUCLEAR PIGULATORY con!ISSIC1!
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l ) Ct.vf i.YCt L(4 Karl R. Co11cr, Assistant Directer for 0;crnting Reactors Division of 0persting'T.cactors Attacheent:
- Chr.nges.to ti:c Technical 1pecifications
' Date of Issuance:
May 19, 1976 t
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. ATTACHMElff TO LICENSE' AMENDMENT NO. 4 FACILITY OPERATING LICENSE NO. DPR-54 DOCKET NO. 50-312 Revise Appendix A as follove:
Remove Panes _
Insert Pages 3-1 & 3-2 3-1 & 3-2
.3-15 & 3-15a 3-15 & 3-15a
'3 3-22 3 3-22 3 3-33c 3 3-33b Figure 3.5.2-1 Figure 3.5.2-1 Figure 3.5.2-2 Figure 3.5.2-2 Figure 3.5.2-3 Figure 3.5.2-3 4 4-28 4 4-28 Figure 6.2-1 Figure 6.2-1 The changed areas on the revised pages are shown by marginal lines.
Pages 3-15a, 3-22, 3-31, 4-25, and 4-28 are unchanged and are included for convenience only.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.'
LIMITING CONDITIONS FOR OPERATION I
3.1 REACTOR COOLANI SUIDI Applicability Applies to the operating status of the reactor coolant system.
_ Objective To specify those limiting conditions for operation of the reactor coolant sys-tem which must be met to ensure safe reactor operations.
3.1.1 OPERATIONAL COMPO!MS Specification 3.1.1.1 Reactor Coolant Pumps A.
Pump combinations permissible for given power levels shall be as shown in specification tabre 2.3-1.
3.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
C.
Operation with two pumpe shall be limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30 day period.
i 3.1.1.2 Steam Generator A.
One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F.
3.1.1.3 Pressurizer Safety Valves A.
The reactor shall not remain critical unless both pressurizer code safety valves are operable.
B.-
When the reactor is suberitical, at-least one pressuriser code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accord-ance.with ASME Boiler and Pressure' Vessel Code,Section III.
Bases A reactor coolant pump or decay heat removal pump is required to be in opera-
, tion before the boron ' concentration is reduced by dilution with makeup water.
Either pump will provide mixing which will prevent. sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump will.circulat9 the equivalent of the reactor coolant system volume in one half hour or~less. W h % e16o.4' 3-l'
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- RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The decay heat removal system suction piping is designed for 300 F and 300 psig.; thus, the system can v
ecay heat when the reactor coolant system is below this temperature One pressurizer code safety valve is capable of preventing overpressurization when the. reactor is not critical since its relieving capacity is greater than
' that required by the sum of the available heat sources which are pump energy,
-pressurizer heaters,' and. reactor decay heat. (4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief :apabilities. The code safety valves prevent overpres-sure for rod withdrawal accidents. (5) The pressurizer code safety valve lif t set point shall be set at 2500 psig +1 percent allowance for error and each valve shall be capable of relieving 345,000 lb/h of saturated steam at a pressure not greater enan 3 percent above the set pressure.
Two pump operation is limited until further ECCS analysis is performed.
i RE"ERENCES (1) FSAR tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6 (2) PSAR paragraph 9.5.2.2 and 10.2.2 (3) FSAR. paragraph 4.2.5 (4) FSAR paragraph'4.3.8.4 and 4.2.4' (5) FSAR paragraph 4.3.6 and 14.1.7.2.3 i
a
- Amendment.No. 4L
~3-2
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
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Limiting Conditions for Operation 3.1.7 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Specification The moderator temperature coefficient shall not be positive at power levels above 95 percent of rated power.
Bases A non-positive moderator coefficient at power le.vels above 95 percent of rated power is specified such that the maximum clad temperatures will not exceed.the Final Acceptance Criteria based on LOCA analyses.
Below 95 with a positive moderator temperature coefficient of +0.9 x 10-4 perce Ak/k/F corrected to 95 percent of rated pcwer. All other accident analyses as reported in the FSAR have been performed for a range of moderator temperature coefficients including +0.9 x 10-4 ok/k/F.
The experimental value of the' moderator coefficient will be corrected to obtain the hot full power moderator coefficient.
When the hot zero power value is corrected to obtain the 95 percent power value, the following corrections will be applied:
1.
Uncertainty in isothermal measurement - The measured moderator temperature coefficient will contain uncertainty owing to the AT -
of the base and perturbed conditions and the uncertainty in the reactivity measurement.
Proper corrections will be added for these conditions to provide a conservative moderator coefficient.
2.
Doppler contribution at hot zero power - During measurement of the isothermal moderator coefficient at hot zero power, the fuel tem-perature will increase by the same amount as the moderator.
The measured temperature coefficient must therefore be increased to obtain a pure _ moderator temperature coefficient.
.1.
Moderator temperature change - The hot zero-power measurement must be corrected for the_ difference in water temperature at zero power (532 F) and 15 percent power (582 F).
Above this power, the average moderator temperature remains 582 F.
4.
Fuel temperature interaction (power effect) - The moderator coef-ficient must be adjusted to' account for the interaction of an average moderator temperature with increasing fuel temperatures 3-15 Amendment No. 4-
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RANCHO S'ECO UNIT 1 TECHNICAL SPECIFICATIONS Lir.iting Conditions for Operation as power increases. Adjus t the me'erator coefficient at 15 percent power to the coefficient at any p eer Ikvel above 15 percent.
.5..
Dissolved baron concentration
""his correction is for any difference in boron concentration between zero and full power.
Since the moder-ator coefficient is more positive for greater amounts of dissolved boron, the sign of the correction depends on whether boron is added
.or removed.
'6.
Contro1 rod insertion - This correction is for the difference.in
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moderator coefficients between an unrodded and rodded core.
7.
Isothermal to distributed temperatures - The correction for spatially 4
distributed woderator temperature effects has been found to be insignificant. Therefore, correction for distributed effects is not required.'
. REFERENCES (1) FSAR, subsections 14.1 and 14.2 (2) FSAR, paragraph 3.2.2.1.5.D l'
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1 3-15a j
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'c RANCH 3 SECO UNIT 1
' TECHNICAL SPECIFICATIONS Limiting Conditions for Operation
~3.3
' EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING, AND REACTOR
-BUILDING SPRAY SYSTEMS Applicability Applies to the emergency core cooling, Rea*ctor Building emergency cooling, and Reactor Building spray systems.
Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, Reactor Building emergency cooling, and Reactor Building spray systems.
Specification 3.3.1
_The reactor shall not remain critical, unless the following conditions
-are met:
A.
Injection System 1.
The borated water storage tank _shall contain a minimum of 390,000 gallons of water having a minimum concentration of 1,800 ppm boron at a temperature not less than 40 F.
The manual valves on the discharge line from the borated water storage tank shall be locked open.
2.
Two out of three high pressure injection pumps shall be operable.
3.
Two safety features actuated decay heat removal pumps shall be operable.
4.-
Both decay heat removal coolers shall be operabic.
5.
Two BWST level instrument channels shall be' operable.
16.
The Reactor Buiiding emergency sump isolation valve shall be either manually or remote-manually operable.
7.
One of the-two BWST isolation valves shall be open (SFV 25063 or SFV 25004)
This valve'may be closed during the quarterly valve test specified in the Specifications 4.5.1.2B and 4.5.2.2B.
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- B.
Core Flooding System 1.
Thetwo.corgfloodingtanksshslleachcontain
'1040 130.t of borated water at 600 +25 psig.
f 2._
. Core floading tank boron concentration shall not be less than 1,800 pra boron.
3.-
The electrically operated dist harge valves _ f ri.n t he u.re flood _ tanks shall be open. The breakers shall.he open and so tagged.
Amendment.No. 4l 3-19
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RANCHO SECO UNIT'l TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 4.
Two core ficod tank pressure instrument channels shall be operable (one per tank minimum).
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5.
The electrically operated vent valves (HY-26511and*HV-26512) from the core flood tanks shall be closed. The breakers shall be open and so tagged except during normal venting operations.
C.
Reactor Building spray system and Reactor Building emergency cooling system.
The following combination of system components must be operabis:
1.
Two Readtor Building spraypumps and their associated spray headers with a minimum of 32 percent NaOH solution in the spray additive tanks and,-
2.
A minimum level of 78 inches of solution shall be available in,each spray addi tive tank.
3.
Four emergency cooling units, two with charcoal filter units.
r D.
Nuclear service cooling and raw water cooling system.
i.
Two nuclear service cooling water (NSCW) pumps and raw water cooling pumps are operable.
2.
The manual valves in the NSCW suction and discharge of each operable Reactor Building emergency cooling unit and at the suction of each NSCW pump are locked open.
3 The manual valves in the suction-and discharge lines of all operable heat exchangers served by the nuclear service raw' water system are locked in their throttled or open position.
E.
Safety features valves and interlocks associated with each of the above systems are operable.
Inoperable valves shall be placed in the safety features position.
3 3.2-Maintenance shall be allowed during power operation on any component (s) in the high pressure, low pressure, nuclear service cooling and raw water cooling, Reactor Building spray, or Reactor Building emergency cooling systems, the core flooding system pressure Instrument channels or BWST leveI channels, which will not degrade safety features system 14 or B below,the level ~of performance with the single subsystem removed from service.
In the context of this specification, a. safety features system consists of the following subsystems: high pressure Injection, low pressure injection, Reactor Building emergency air cooling, Reactor Building spray, diesel generator, nuclear se-vice cooling water and nuclear service raw water.
If the system being repaired is not restored to meet the requirements of specification 3.3.I within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a hot shutdown
' condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
if the requirements of specification 3 3.1 are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shgil be placed in a cold shutdown condition within 24' hours.
Amendment No.=4 3-20 i
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.. RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.3.3 Prior to-initiating maintenance on,any of the components, the dupli-
- cate (redundant) components shall be tested to assure operability,
.with the component on which maintenance is being performed removed from service.'
Bases-4
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The requirements of Specification 3.3.1 assure that, before the reactor can be made critical, adequate safety features are operable. Two high pressure injection pumps and two decay heat removal pumps are specified.. However, only
.one of'each'is necessary to supply emergency coolant to the reactor in the event of a loss-of-coolant accident. Both core flooding tanks are required
- as m' single core flood tank has' insufficient inventory to reflood the core.(1) 1 The borated water storage tank'is-used for two purposes:
A.
As a supply of bo' rated water for accident conditions.
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B.
-As a' supply of borated water-for flooding ~the fuel transfer canal during refueling operation.(2) 390,000 gallons" of borated water,are supplier. for emergency core cooling and Reactor Building spray in the event of a loss-of-core coolant accident.
This amount - fulfills; requirements for emergency core cooling. The borated.
water storage. tank minimum volume of 390,00') gallons is based on refueling vol-ume-requirements. LHeaters maintain the bo.ated water supply at a temperature to prevent freezing. The boron concentratica is set at the amount of boron
- required to maintain'the core 1 percent suberitical at 70 F without.any. con-trol rods in the
- core. This concentrr. tion is 1585 ppa boron while the mini-mun value specified in the tanks is 1,800' ppa boron.
The requirement that one.BWST. isolation valve shall be open assures a static head to t'.ie injection pump not lined up to the makeup tank.
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' The post accident Reactor Building cooling may -be accomplished by two spray units' or by a. combination of two emergency cooling units and one' spray unit.
The specified requirements assure that the required post accident components
-are available.
'The spray system utilizes common suction lines with the decay. heat renoval
- sys tem. _~ If a. single train of equipment is removed from either system, the other train 'must'.be assured to be ' operable in each system.
When/the reactor.is critical, maintenance is allowed per Spec *?ication 3.3.2 provided tegoirements'in Specification.3.3.3 are met which r.ssure
,' operability.of theLduplicate' components.. Operability of inc specified com-
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ponents'shall be based.on:the results of testing as required.by Tech:ical
)
-Specification 4.5. ~Tho'. maintenance period-of up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable if.
- tt,tc. operability ~of equipment redundant to that removed f rer. service is' der:en-c atrat ad --inmediately - su'osequent l to' removal.. The basis of.acceptahility is a.
low: likelihood ofifailure within-48 hours following such demonstration.
AmandaantNod4 3 ~
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Ccnditions for Ope)ation In the event that the need for emergency core cooling should occur, function-ing of one train. (one high pressure injection pump, one decay heat removal pump, and both core flooding tanks) uill protect the core and in the event of
-a main coolant loop severence', limit -the peak clad temperature to less than 2,300-F and the metal-water reaction to less than 1 percent of the clad.
The nuclear service cooling water system consists of two independent, full capacity, 100 percent redundant systems, to ensure continuous heat tiemoval.(3)
P.EFERENCES (1) FSAR, paragraph 6.2.1 (2) FSAR, paragraph 9.5.2 (3) FSAR, paragraph 9.4.1 3-22
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1 RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS
' Limiting Conditions for Operation 3.5.2 CONTROL ROD GROUP AND POWER DISTRIBUTION LIMITS
' Applicability This specification applies to power distribution and operation of control rods during power operation.
Objective To assure was acceptable core power distribution during power operation, to set a limit on potential. reactivity insertion from a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip.
Spycification 3.5.4.1 The available shutdown margin shall be not less than 1 percent ak/k with the highest worth control red fully withdrawn.
3.5.2.2 Operation with inoperable rods:
A.
Operation with more than one inoperable rod as defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall-not be permitted.
B.-
If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification paragraph 4.7.1.1-and 4.7.1.3; an evaluation shall be initiated immediately to verify the existenc.e of 1 percent ak/k hot shutdown margin. Boration may be initiated to increase the available rod worth either to compensate for the worth of the inoperable rod o'r until' the regulating banks are fully with-drawn,'whichev,er occurs first.
C.
If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determitted that a 1 percent Ak/k hot shutdown margin exists combin?.ng the worth of the inoperable rod with' each of the ?ther rods, the reactor shall be brought to the hot standby condition until this margin is established.
D.
' Following the de' termination of an inoper rod as defined in Specification 4.7.-1, all rods shall be ex,.eised by a movement until indication is noted but not exceeding 2 inches within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod probica is solved.
E.
If a control rod in the rcquiating or safety rod groups is declared inoperable per 4.7.1.2, power shal; be reduced to 60 percent of the thermal power allowable for the reactor
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coolant pump combination.
3-31 vem
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RANCHO SECO UNIT 1 t
. TECHNICAL SPECIFICATIONS l
t Limiting Conditions for Operation F. -
If a control rod in the regulating or axial power shaping 1
groups is declared inoperable per Specification 4.7.1.2, oper-ation above _60% of. rated power may continue provided the rods 4
in the group are ' positioned such that the rod that was declared inoperable is maintained within' allowable. group average posi-J tion limits of Specification 4.7.1.2 -and the withdrawal limits 3;
of Specification 3.5.2.5.c.
3.5;2.3 The worth of a single. inserted -control rod shall not exceed 0.65 per-
{
cent Ak/k at rated power or 1.0 percent ak/k at hot zero power except for physics testing.when the requirement of Specification 3.1.8 shall apply.
3.5.2.4 Quadrant tilt:
A.'
Whenever the quadrant power tilt exceeds 4 percent, except for physics tests, the quadrant tilt shall be reduced to less than
. 4 percent within two hours or the following actions shall be takent (1)
If'four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of maximum allowable power for each 1 percent tilt in excess of 4 percent. The allowable thermal power is defined by l
Figure. 3.5.2-1 where the power. level cut off may apply during transient xenon operation.
(2)-
If less than four reactor coolant pumps are'in operation, the allowable thermal power shall be reduced by 2 percent of maximum allowable power for each 1 percent tilt below the power allowable for the reactor coolant pump com-bination.
(3); Except as provided in 3.5.2.4.b, the reactor shall be Lbrought ' to the hot shutdown condition within four hours if the quadrant' tilt is not reduced co less than 4 per-cent after 24 houro.
(4)
The power range high flux set point shall be reduced 2 percent of the maximum allowable flux for the RC pump combination for each 1 percent tilt in excess of 4 percent.
'B._
If the' quadrant tilt'ex'ceeds 4 percent and there is simulta '
neous indication of a misaligned contro1 ~ rod per Specification 3.5.2.2, reactor ' operation may continue, provided power is reduced to 60 percent of. the thermal power allowable for the reactor' coolant pump combination.
C.-
Extept for physics tests if[ quadrant tilt exceeds 9 percent, the reactor.shall be brought ~to the hot shutdown condition
? within four hours.
- Amendment
- No. 3,J4.
3-32 m
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RANCHO SECO UNIT.-1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation D.1 Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4a(3) or 3.5.2.4c above, subsequent reactor operation is permitted for the purpose of measurement, testing and corree-tive action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combina-
~ tion are restricted by a reduction of 2 percent of maximum
. allowable power for.each 1 percent tilt.
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E.
. Quadrant Fower tilt shall be monitored on a minimum frequency.
of once every two hours during power operation above 15 percent 1
of rated power.
3.5.2.5 Control Rod Positions A.-
Technical Specification 3.1.3.5 (sefety rod withdrawal) does not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical. Specification 3.5.2.,2.
B.
Operating rod group overlap shall be 25 percent 15 percent between two sequential. groups, except for physics tests.
C.
Except for physics test or exercising e>ntrol rods, che con-trol. rod. withdrawal limits are specific 1 on Figure 3.3.2-1.
If control rod position limits are exceeded, corrective measures -shall be taken immediately to achieve an acceptable a
control rod position. Acceptable control rod positions shall i
then be attained within two hours.
- D.
Except for physics test, power shall not be increased above the powet leve1~ cut-of f 'of 92 percent of the maximum allowable power level unless one of the following conditions is satis-fled:
(1)-
Xenon reactivity is witnin 10 percent of the equiliabrium
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value for operation at the maximum allowable power level and asymptotically approaching stability.
(2)
Except for. xenon free start-up, when 3.5.2.!D(1) applies.
the reactor has operated within a range of 87 to 92 per-i cent'of -the maximum allowable power for a period exceed-ins 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control mode.
l 3.5.2.6 ~ Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power opeation above 40 percent ' rated power.. Except for physics test, imbalance shall be maintained l
. within the envelope defined by Figure -3.5.2-2.
If the imbalance l
1s not within the envelope defined by Figure 3.5.2-2, cc ective j
measures shall be taken to achieve an acceptable imbalat a.
If L
an acceptable imbalance is not achieved within two hours. reactor power shall be reduced until imbalance limits aremet.
Amendment-No~$'4?
3-33 f
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation.
3.5.2.7 The control' rod drive patch panels shall be locked at all times with 2
limited access to be authorized by the superintendent or his designated irepresentative.
The power-imbalance envelope defined in Figure 3.5.2-2 is based on LOCA analyses which have defined the maximum linear heat rate (see' Figure 3.5.2-3) such that the ' maximum clad _ temperature will not exceed the Final Acceptance Criteria.
Operation outside of the power imbalance envelope alone does not constitute a; situation that would cause tho Final Acceptance Criteria to be
+
exceeded should a LOCA occur.
The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the position limits as defined by Figure 3.5.2-1 and if a l
4 percent. quadrant power tilt exisi.s. -
by application of t Additional conservatism is introduced A.
Nuclear uncertainty' factors.
B.
'Ihermal calibration. uncertainty. -
C.
Fuel densification effects.
i D..
Hot rod manufacturing tolerance factori,
t The 25, percent overlap between ' successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
rods are arranged in groups
- defined as follows:
Control i
_ Function 1
Safety 2
i Safety 3
safety
'4 Safety 5
Regulating 6-7 Regulating Regulating 8
APSR (axial power shaping bank)
Control rod groups are withdrawn in sequence beginning with Group 1.. Group 5 is overlapped 25 percent with Groups 6 and 7, which operate in parallel.
The
. normal ~ position at power is 'for Groups 6 and 7 to be partially inserted.
- The minimum available rod worth provides for' achieving' hot shutdown by' reactor trip at'any t me
.out position.1) assuming the highest worth control rod remains in the full t
Inserted rod groups during power operation will not contain single rod worths-greater L than 0.65 percent. Ak/k.
This value has been_shown safety analysis of the hypothetical rod ejection accident.(to be safe by the 4)- A single
~
! inserted control rod worth of 1.0 percent Ak/k at beginning of life, hot, sero. power would result-in the:same transient peak thermal power and therefore E the ' sue environmental consequences 'as a 0.65 percent Ak/k ejected rod worth ct rated power.
p Amendment 1No.3,[.4' ic F33a i
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant power. til t given in Technical Specifications, Section 1.6.
These
= limits in conjunction with the control rod position limits in Specification 3.5.2.5c ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.
The quadrant tilt and axial imbalance monitoring in Sepcifications 3.5.2.4 and 3.5.2.6, respectively, normally will be performed in the process computer.
The two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.
Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without spe,cification viola tion. Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.
Operating restrictions are included in Technical Specifications 3.5.2.5d(1) and 3.5.2.5d(2) to prevent excessive power peaking by transient xenon. The xenon reactivity must either be beyend the "undershoot" region and asympto-tically approaching its equilibrium value at rated power or the reactor must be operated in the range of 87% to 92% of the maximum allowable power for a l
period exceeding two hours in the soluable poison control mode so that the tranetent peak is burned out at a lower power level.
During physics testing, additional safety margins are provided by administra-tively setting special reactor protection system limitations. During the power ascension testing program, the following high flux trip. settings will be set prior to increasing power to the next plateau:
Test Plateau Lovel %
- Overpower Trip %
0
<5 15 50 40 50 75 95 90 100 100 1
05.5 REFERENCES
(1) FSAR, Paragraph 3.2.2.1.2 (2) PSAR, Paragraph.14.2.2,4 Amendment No. J, 4 3-33b
200,102 100 182.4.102 -
i 90
' POWER LEVEL CUT 0FF 80 70 RESTRICTED REGION 60 5
g 50 Z
122.3.43 PERWISSIBLE O
OPERATING y
REGION d'
30 20 80,15 10 49.o O
l I
l I
I I
l' l
I O
20 40 _ 60 80 100 120 140 160 180 200 Rod index, 5 Withdrawn 0
25 50 75 100 1
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O 25 50 75 100 Group 6 & 7 I
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- 1. Rod incex is the percentage su:2 of the withdrawal of control rod groups 5.6 and 7.
- 2. The restrictions on rod position are in effect after i
100 EFPD.
ROD POSITION LIMITS Amendment No.' 3,4 Figure-3.5.2-1 i
120 1
RESTRICTED REGION 110
-15.3,102
+10,2,102 100
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-23.0.82
+13.6,8G 80 g
PERMISSIBLE 70 OPERATING
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+40
+60 Core imbalance, 5
- 1. The limits'are in effect after 100 E'FPD.
Amendment No. 3, 4 OPERATIONAL POWER IMBALANCE ENVELOPE Figure 3.5.2-2
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18 N
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10 12 Axial Location of Peak Power From Bottom of Core, ft LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE Figure 3.5.2-3 l
i Amendment No. 4 1
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RANCHO SECO UNIT I TECHNICAL $PECIFICATIONS Surveillance Standards 4.4.2 STRUCTURAL INTEGRITY Applicability Applies to the structural Integrity of. the Reactor Building.
Objective To define the inservice surveillance program for the Reactor Building.
Speelfication 4.4.2.1 Tendon Surveillance An Inspection as described below for lift-off measurements, strand survelliance and anchorage surveillance shall be performed I, 2 and.3 years after-the initial containment integrity test and every 5 years thereafter.
4.4.2.2 Lift-Off Measurements Lif t-off measurements of the prestress force shall be made on the following:-
A.
Six dome. tendons, 3 normal and 3 modified,with one of each in each 60* group.
B.
Six vertical tendons, 3 normal and 3 modified.
C.
Six hoop tendons, 3 normal and 3 modified; modified hoop ten-dons are provided wish shim stack at each anchorage to allow detensioning.
The lift-of f readings sha11 not be less than the values predicted for the particuiar came the inspection is made. These predicted values shall be based upon the final Jacking forces corrected for recorded seating losses and calculated losses due to concrete creep and shrinkage and pre-stressing steel relaxation.
Af ter the initial lift-off readings have been taken on the modified I
hoop-tendons they shall be Jacked to a force of.8f's and then com-l pletely detensioned to inspect for broken cr damaged strands.
Strand continuity will be verified during retensioning by comparison of pre-dicted and observed elongations and' forces.
.All other surveillance tendons will be Jacked only to. observe and record lift-off.
l 4-21 p
Amendment No.-4 "CP W
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RANCHO SECO UNIT I
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' TECHNICAL SPECIFICATIONS Surveillance Standards
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Specification (Continued) 4.4.2.3.
Strand Surveillance-
-l At each inspection one strand shall be removed from one of the modi-fled tendons in the dome, hoop and vertical directions. Portions of these strands'will be tensile tested and an examination will be l
conducted over their entire length to determine if evidence of corrosion or other deleterious effects are present..At each j
. successive inspection the sample selection will be rotated among the 9. modified. tendons.
Should the inspection of one of the strands reveal any. significant corrosion (pitting or loss of area), further inspection of the _
other two modified tendons in that directional group will be made to determine the extent of the corrosion and its significance to the load carrying capability of the structure.
If significant corrosion -is observed at any position in a strand, a tensile test will be made on a specimen, including the length with corrosion representative of the maximum observed.
. Tensile tests will be made on a minimum of three specimens taken from the ends and middle.of each of the three strands. These specimens will be tested to the requirements of ASTM 416 for 270,000_ psi ultimate strength strand.
4.4.2.4 Anchornae Surveillance The' tendon anchorage assembly' hardware of all tendons inspected will be visually checked. The surrounding concrete will also be checked visually for indications of abnormal material behavior.
The anchor heads and visible portions of wedges of all tendons inspected will be checked for grease coverage and evidence of corrosion.
Saari of grease will be 'taken from the 3 tendons that have strands removed under strand surveillance above. ~These grease samples will be checked by laboratory analysis for acceptance per the requirements of the grease specification included in the C12.1A supply specification for the tendon system.
Additionally, the following will be checked on the modified hoop Ltendons at each inspection:
A.-
The shims, trumpet and tendon in the anchorage area will be inspected for grease coverage and signs of corrosion.
. Amendment No. 4
.4-22:
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RANCHO SECO UNIT 1 i
TECHNICAL SPECIPICATIONS i
Surveillance Standards SpectfIcatlob (Continued)
B.
The grease _ coverage will be noted along with the temperature to accumulate a record of grease variation versus temperature.
i
. 4.4.2.5 Repores A report covering the results of each inspection will be prepared, reviewed by the District's Generation Engineering Department and filed with the plant quality assurance records. If any significant or critical deterioration is noted by this inspection, it will be i
reported to the NRC as a reportable occurrence in accordance with Technical Specification 6.9-1.
Should this be necessary, the initial I
report may be made within 10 days of the completion of the tests and the detailed report any follow within 90 days of the completion of
.the tests.
I 4.4.2.6 Liner Plate Surveillance The liner plate will be examined prior to the initial pressure test in accessible areas to determine the following:
l A.
Loca'tlon of areas which have inward deformations. The magnitude of the inward deformations shall be measured and recorded.
These areas'shall be permanently marked for future reference and the inward deformations shall be measured betneen the angle stifferers which are on 15-inch centers. The measurements shall be accurate to 0.01 inch. Temperature readings shall be obtained on both the liner plate and outside containment wall j
at the locations where inward deformations occur.
8.
Locations of areas.having strain concentrations by visual examination with emphasis on the condition of the liner surface. The location of these areas shall be recorded.
4.4.2.6.2 Shortly after the Initial pressure test and approximately one year af ter initial start-up, a re-examination of the' areas located in paragraph 4.4.2.6.1.A shall be made. Measurements of the inward deformations and observations of any strain concentrations shall be made.
4.4.2.6.3 If the difference in the measured inward deformations exceeds 0.25 inch (for a particular location) and/or changes in strain concen-tration exist, and investigation shall be made. The investigation will determine any necessary corrective action.
4.4.2.6.4 The surveillance program shall be discontinued after the one year initial start-up Inspection if no corrective action was needed.
if corrective action is required, the frequency of inspection for a continued surveillance program shall be determined.
Amendmen't No. 4 4-23 y
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t RANCHO SECO 4JNIT 1 TECHNICAi, SPEC,IF1 CATIONS Surveillance Standards Ba ses Provisions have been made for an in-service surveillance program, covering the first five years of the life of the unit, intended to provide suf ficient evidence 4
to maintain confidence that the integrity of.the reactor building is being pre-served.. This program consiste of tendon, tendon schorage, and liner plate surveillance.
To accomplish these programs, two separate sets of nine tendons each are used.
- Each of the sets consists of three horizontal tendons, three vertical tendons and three dome tendons. The locations of these 18 tendons are shown in Figure 5A-21.
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' In its normal configuration, the VSL wedge anchored ' strand tendon system can-l not be detensioned without destroying the tendon. The anchorages of three
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hoop tendons have been modified by the addition of shims to permit them to be detensioned. The chias are placed between the bearing plate and the anchor head prior to initial tensioning and are of a total length at least equal to l
the tendon elongation. During surveillance, these shims are removed in incre-4 ments until the tendon is detensioned. Modified dome and vertical tendon have additional length extending beyond the anchor head to facilitate removal of a j'
corrosion surveillance strand.
Strand continuity cannot be checked by pulling each strand to observe its movement at the opposite end since the wedges are held in the anchor head by
- a residual clamping force af ter the tendon is completely detensioned. The.
wedges should not he dislodged since it is not advisable to regrip the strand in the same place.-
The inspection during this initial five year period of at least one strand from each of the 'nine corrosion surveillancq tendons is considered sufficient representation to detect the presence of any wide spread tendon corrosion or i
pitting conditions in the structure. This program will be subject to review and revision as.werranted based on studies and on results obtained for this and other prestrested concrete react'or buildings during this period of time.
REFERENCE FSAR Paragraph 5.2.5.3 i
Amendmene No.[4 4-24 6
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.4.3-HYDROGEN PURGE SYSTEM Applicability Applies to testing Reactor-Building hydrogen purge system.
Objective-
. To. verify that this systen and components are operable.
Specification 4.4.3.1 Operating Tests An in-place. system test shall be performed during each refueling -
4 interval..These tests shall consist of visual inspection and a flow measurement using the installed flow instruments. Flow shall be design flow or higher., Blower motors shall be operated contin-uously. for at least one hour and valves shall be proven operable.
- 4. 4. 3.,2 H2 Detector Test The hydrogen concentration analyzer,shall be cal _brated
-yearly.
Bases
.The hydrogen purge system is composed of. permanently installed, independent, centrifugal enhaust blowers, portions of the Reactor Building radiation moni-toring and sampling system, and the waste gas system. This controlled purge system is completely independent of the larger purge system employed, as required, during normal plant operation prior to building entry. Since this system is not normally operated, a periodic test is required to insure its
- operability when needed. During this test the system will be inspected for suchithings as water, oil, or oth'er foreign material; gasket deterioration;
-and. unusual or excessive noise or vibration when the blowers are running.
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TECHNICAL SPECIFICATIONS Surveillance Standards 4.5: EMERGENCY CORE C00 lib'. AND REACTOR BUILDIND COOLING SYSTEM PERIODIC TESTING
%.5.1' EMERGENCY CORF COOLING,SYSTDI Applicability:
Applies to' periodic testing requirement for emergency core cooling systems.
Objective
~ To verify that the emergency core cooling systems are ope.rable.
Specification 4.' 5.1.1 System Test's A.
-High Pressure. Injection L
During each refueling interval, a makeup and purification system test shall be conducted to demonstrate the system.is operable for high pressure injection. A manual trip signal will be applied to demonstrate actuation of the makeup and purification for emergency core cooling operation.
2.
-The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal; all appropriate pump breakers shall have opened or closed and all valves shall have completed their travel.
3.-
The high pressure injection pump casings shall be vented monthly and prior to any ECCS flow tests.
- B.-
. Low Pressure Injection 1.
During each refueling interval a decay heat removal system test shall be conducted to demonstrate the system
-is operable for low pressure injection. The test shall be performed in accordance with the procedure summarized below:-
A manual _. trip signal will be applied to demon-a.
strate ' actuation of the decay heat ' removal-system a
for emergency co're cooling operation.
.b.
. Verification of the safety features function of the
. nuclear service cooling water. system and nuclear
,s
-service raw water s'ystem which supplies cooling water to the decay heat removal coolers shall be made to demonstrate operability of the coolers.
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l RANCHO SECO UNIT 1
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TECHNICAL SPECIFICATIONS Surveillance Standards 2.
The test will be considered satisfactory if control board indication verifies.that all components have responded to the actuation signal and all appropriate pump breakers shall have opened or closed, and all valves have completed their travel.
I 3.
Decay heat pump castags shall be vented monthly and prior to any ECCS flow tests.
C.
Core Floodidg System 1.
During each refueling interval, a core flooding system
-test shall be conducted to demonstrate proper operation of the system. During depressurization of the reactor coolant system, verification shall be made that the check valves in the core flooding tank discharge lines operate.
2.
The test will be considered satisfactory if control board indication of core flood tank level verifies that all check valves have opened.
4.5.1.2 Components Tests A.
Pumps At least quarterly, the high prpssure, makeup, and decay heat removal pumps shall be started and operated to verify operation.
B.. Valves -- Power Operated At least quarterly each safety features valve in the emergency core cooling systems and each safety features valve associated with emergency core cooling in,the decay heat. removal system shall be tested to verify operability.
C.
Nuclear Service Cooling and Raw Water System
~At' least quarterly, the' nuclear service co'oling and raw water system shall be operated to verify performance.
D.
Acceptance 1.
Acceptable performance for the high pressure injection pumps shall be that the pump starts.and operates for 15 minutes di-scharging through the miniflow and the dis-
. charge. pressure indicates, flow is within 10 percent of
_t'he initial level of performance and at least equal to the. minimum design flow rate.
2.
Acceptable performance for the decay heat pumps shall be that the pump starts and. operates for 15 minutes discharg-ing through-the test flow path and the discharge pressure and flow are within 10 percent of the init'ial level of performance and at least equal to the minimum design flow rate.1 4-27 A===A==1t No. 4
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rN RANCHO SECO UNIT 1
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' TECHNICAL SPECIFICATIONS Surveillance Standards 3.
_ Acceptable performance for the nuclear service cooling water and raw water _ pumps shall be that the pump starts and operates for 15' minutes, at the design flow rate with the required differential pressure.
-4.
The acceptable performance of each power-operated valve
. in the emergency core cooling system will be that motion is indicated upon actuation by appropriate signals.
Bases The emergency core cooling systems are the principal reactor safeguards in the event of a loss-of-coolant accident. The remeval of heat from the core pro-vided by these systems is designed to limit core damage.
The decay heat removal pumps are tested singularly for operability by opening the-borated water storage tank outlet valves and the test line valves to the borated water storsge tank. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank through a test'line.
With the reactor shutdown, the check valves in.each core flooding line are checked for operability. by reducing the rear,cor coolant ' system pressure until the indicated level in the core flood tanks, verify the check valves have opened.
REFERENCES FSAR subsection,6.2.
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4-28 4
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BOARD OF DIRECTORS SF.CRETARY TREASURER ATTORNEY ACCOUNTANT INDEPENDENT CONSULTANTS MANAGDIENT PERIODICSAI;LTY AUDIT GENERAL MANAGER sal:LTY RI'. VIEW OP RANCilO SECO OPERATIONS COMMITTEE I
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l CONSUMER PERSONNEL RELATIONS LEGAL STAFF DEPARTMENT DEPARTMENT I
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ACM*
AGM*
AGM*
CONTROLLER OPERATIONS CHIEF ENGINEER SYSTEMS & PROCEDURES STAFF ASSURANCE QUALITY
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INTERNAL AUDITING ACCOUNTING DISTRIBUTION ENGINEERING CONSTRUCTION
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DEPARTMENT DEPARTMENT DEPARTMENT 4
DATA DISTRIBUTION HYDRO PROCESSING OPERATIONS PLANNING DEPARTMENT DEPARTMENT DEPARTMENT GENERAL DISTRI8tJTION NUCLEAR SERVICES OPERATION OPERATIONS DEPARTMENT DEPARTMENT DEPARTMENT PURCHASES CUSTOMER GENERATION
& STORES SERVICES ENGINEERING
- DEPARTMENT DEPARTMENT DEPARTMENT l
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- ACM. Assistant General Maasser FIGURE 6.21 SMUD ORGANIZATION CHART
, RANCHO SECO UNIT 1-
- TECHNICAL SPECIFICATIONS -
Amendment No. 41
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