ML18152A479

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Proposed Tech Specs Revising Operability Requirements for Individual Rod Position Indicating Sys
ML18152A479
Person / Time
Site: Surry  Dominion icon.png
Issue date: 08/05/1988
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18152A480 List:
References
NUDOCS 8808150022
Download: ML18152A479 (23)
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Text

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ATTACHMENT I Proposed Technical Specification Changes 8808150022 880805 PDR ADOCK 05000280 p PDC

l,,' TS 3.12-8

~T and Overtemperature ~T trip settings shall be reduced by the equivalent of 2% power for ever 1% quadrant to average power tilt.

C. Inoperable Control Rods

1. A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaTignecfrrom rts group step oemana position by more than +/-24 steps during the "Thermal Soak" period, as defined in Section 3.12.E.l.b, or +/-12 steps otherwise during power operation. No tolerance limit is required in the shutdown modes, but a rod shall be considered inoperable if the rod position indicators . do not
  • verify rod movement upon demand. Add it i ona 11 y, a full-length control rod shall be considered inoperable if its rod drop time is greater than 2.4 seconds to dashpot entry.
2. No more than one inoperable control rod assembly shall be permitted when the reactor is critical.
3. If more than one control
  • rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanism, (i.e. programming circuitry), the provisions of Specifications 3.12.C.l and 3.12.C.2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs. In the event the affected assemblies cannot be returned to service within this specified period, the reactor will be brought to hot shutdown conditions.
4. The provisions of Specifications 3.12.C.l and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned.
5. Power operation may continue with one rod inoperable provided that within one hour either:
a. the rod is no longer inoperable as defined in Speci-fication 3.12.C.l, or

L-e 3,12-10

5. If power has been reduced in accordance with Specification 3.12.C.5.b, power may be increased above 75% power provided that:

a) an analysis has been performed to determine the hot channel factors and the resulting allowable power level based on the limits of Specification 3.12.B.1, and b-)~an-eva-luatfon-of-the-ef fec-t-s-of-oper-at-i-ng-at-t-he-----

i ncreased power level on the accident analyses of Table 3.12-1 has been completed.

D. Core Quadrant Power Balance:

1. If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined:
a. Once per day, and
b. After a change in power level greater than 10% or more than 30 inches of control rod motion.
2. The core quadrant power balance shall be determined by one of the following methods:
a. Movable detectors (at least two per quadrant)
b. - Core exit thermocouples (-at least -four per quadrant)

E. Rod Position Indicator Channels

1. Rod Position Indication shall be provided as follows:
a. Above 50% power, the rod position indication system shall be operable and capable of determining the control rod positions to within +/-12 steps of their respective group step demand counter indications.
b. Between 0% and 50% power, the rod position indication system sha 11 be operable and capable of determining the control rod positions to within +/-24 steps of their respective group step demand counter indications for a maximum of one hour out of twenty-four, and to within +/-12 steps otherwise.

During the one-hour "Thermal Soak" period, the step demand counters shall be operable and capable of determining the group demand positions to within +/-2 steps.

TS 3.12-11

c. In hot, intermediate and cold shutdown conditions, the step demand counters shall be operable and capable of determining the group demand positions to within +/-2 steps. At 1east two rod position indicators per group shall be available to verify rod

_________________ mo_v_ement_up.on_demand_. _____________~_,______

2. If a rod position indicator channel is out of service, then:
a. For operation above 50% of rated power, the position of the RCC shall be checked indirectly using the movab 1e i ncore detectors at 1east once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and immediately after any motion of the non-indicating rod exceeding 24 steps, or
b. Reduce Power to less than 50% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. During operations below 50% of rated power, no special monitoring is required.
3. If more than one rod position (RPI) indicator channel per group or two RPI channels per bank are operable, then the requirements of Specification 3.0.1 will be followed.

The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concentration. During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups.

A reactor trip occurring during power operation will place the reactor into the hot shutdown condition. The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient. margins to meet the assumptions used in the accident analysis. In addition, they provide a limit

TS 3.12-13 in service, the effects of malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors. Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assembly 12 feet out of alignment with its bank), operation

--:------ -~ -------~------at---50% -ste-a-dy-state-pmoJer--dn-e-s---mrt-result in exceeding core ---

1imits.

The "Thermal Soak" allowance below 50% power, during which the rod position indication system to 1erance requirement is relaxed, provides time for the system to reach thermal equilibrium. A total of one hour in twenty-four is available for this allowance, which may be a continuous hour or may consist of discrete, shorter intervals. For such a short period of time, a misaligned rod does not pose an unacceptable risk. At these conditions, the rod position indicators should still be used to verify rod movement but not their exact location. The tolerance is tightened after one hour to ensure that the thermal overshoot does not conceal an actual rod misalignment.

The re 1i ance upon the step demand counters at hot and co 1d shutdown conditions shifts the monitoring of rod position from the rod position indication system to the more reliable demand counters when RCS temperature is changing greatly but the core remains subcrit i ca 1 . The step demand counters a1so provide precise group demand positions during the thermal soak period.

The specified control rod assembly drop time is consistent with safety analyses that have been performed.

TS 3.12-13a An inoperable control rod assembly imposes additional demands on the operators. The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemb,lies upon reactor trip.

Two criteria have been chosen as a design basis for fuel performance reTaEec:lto-rfssion gas release, pellet-temperature, and cladding mechanical properties. First, the peak value of fuel centerl i ne temperature must not exceed 4700 ° F. Second, the minimum DNBR* in the core must not be less than the applicable design limit in normal operation or in short term transients.

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TAfilE 4.1-1 MINDDI FREuJEN< :1 FS FOR ClIFXX1 CALIBRAT.[<m AND

'ffS'l' OF INS'.lH1MFNl' (]JANNE[S Cllannel Description Clleck calibrate Test I Remarks

1. Nuclear Power Range s D(l) M(2) 1) Against a heat balance standard Q(3) 2) Signal at 6, T; bistable action R(4) . (~ssive, rod stop, trip)
3) Upper and lower chambers for sy#nnetric offset by means of th~ movable incore detector 4)

I system Neiltron detectors may be ex-e cltlded from Channel calibration

2. Nuclear Intennediate Range (below P-10 setpoint)
  • S R(2) P(l) 1) J level; bistable action (pennissive, rod stop, trip)
2) Neiltron detectors may be ex-clMed from Channel calibration
3. Nuclear Source Range *S R(2) P(l) 1) Bibtable action (alann, trip)

(below P-6 setpoint) 2) Netltron detectors may be exct:luded from Channel calibration

4. *Reactor Coolant .Temperature *S R M(l) 1) O V ~ t u r e 6.T I

M(2) 2) OV~er 4-T

5. Reactor Coolant Flow s R M e
6. Pressurizer Water Level s R M.
7. Pressurizer Pressure (High & I.a-/) s R M
8. 4 KV Voltage and Frequency N.A. ;R M

.1 >-3 9 *. Analog Rod Position *S(l,2) R M(3) 1) With step counters (/)

(4) 2) Ea¢h six inches of rod motion whfim data logger is out of I '

.i:,.

1--'

service

  • I
3) Rocl bottom bistable action 0)
4) N. t*

I ' '

when reactor is rn hot, in ermediate or cold shutdown

  • I

r TABIB 4 .1-1 (amtinued)

MINlMJM FREXXJENCIES FOR CHEll{ 1 CALIBRATIONS AND TEST OF INSIRJMFNl' ClIANNEI.S Cllannel Descriotion OJ.eek calibrate Test *Remarks i

I

10. Rod Position Bank Counters S(1,2) N.A. N.A. 1) Each six inches of rcxi motion Q(3) when data logger is out of I
  • service
2) W~th analog rcxi }X)Sition
3) For the control banks, the ~

benchboard indicators shall be tjlecked against the output of

  • the bank overlap unit.

I

11. steam Generator level s R M
12. Cllarging Flow N.A. R N.A.
13. Residual Heat Removal Pump Flow N.A. R N.A.
14. Boric Acid Tank level *D R N.A.
15. Refueling Water storage Tank s R M level
16. Volmne Control Tank level N.A. R N.A. i
17. Reactor Contairnnent Pressure- *D R M(l) 1) rkoration valve signal and CIS spray signal
18. Boric Acid Control N.A. R N.A. I
19. Contairnnent SUmp level N.A. R N.A.
20. ACCLnnUlator level & Pressure s R N.A.

t-3

(/)

21. Contairnnent Pressure-Vacuum s R N.A.

Pump System .i:,.

i-'

I

22. steam Line Pressure s R M -.J

ATTACHMENT 2 Safety Evaluation

( (.

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,, e I. INTRODUCTION Surry is one of several Westinghouse designed plants which have experienced difficulties with calibration and accuracy of the Individual Rod Position Indication System (IRPIS). The system consists of a set of

-- -1 fne-ar vari abre ~tra:r1sform-er-s --formed--from--pri-mary-and-seconcfar--y- -CO-iJ s---- __ - ~ __ _

alternately stacked on cy1indrical stainless steel tubes. An extension shaft from the rod drive mechanism extends up into the tube and serves as a variable core for the transformer._ Thus, with a constant a.c.

current source applied to the primary side, movement of the rod drive extension shaft changes the primary to secondary coupling of the transformer and produces a secondary voltage that is directly related to rod position.

There are two basic defi ci enci es in the IRPIS performance, however.

First, the instrumentation readout design is based on the assumption that secondary output voltage is a linear function of rod position. In fact, the steady-state calibration curve is an arc-shaped or even an S-shaped curve. This deviation from linearity is normally absorbed by a +/-12 step Technical Specifications allowance for rod misalignment.

The second and more ~erious drawback is that the instrument response is highly temperature sensitive, and there is a transient (nonequilibrium) temperature response associated not only with RCS temperature changes but also with rod motion. On moit plants the IRPIS for each individual rod is calibrated at hot operating temperatures at beginning of cycle. As the reactor is cooled to hot and cold shutdown Page 1

the hot calibration curve becomes inaccurate and may be off by as much*

1 as 60 steps, or over one-quarter of the core h~ight. .

The transient thermal response problem has been charatterized as an 11 overshoot 11 for most plants. In other words, ifa rod is withdrawn, the

~--I-RPI-S -wi-1-1-show-gl'!ea-ter-w-Uhd~awaJ_than-'--a.c.t_ua 1 for a geri od of time, and 1

then as the system returns to thermal equilibrium the indication will settle back to the 11 true 11 calibrated value. Similarly, if the rod is inserted, the IRPIS will initially indicate a greater insertion than actual. Reference 2 indicates that the transient indication error is as high as 25 steps for some plants and tends to be worse near the fully 11 withdrawn position. The thermal soak 11 tim*e, or the time for the IRPIS to reach equilibrium following rod motion, is reported to range between 20 and 45 minutes.

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II. DISCUSSION AND EVALUATION As a result of these difficulties, Virginia Electric and Power Company is proposing a set of changes to the Technical Specifications which shifts the emphasis from the IRPIS to the demand position indication system for

- ---:----- - - --pro v-i d-i ng-de t-a-i 1-ecl- rocl--p os-i-t-i-o n -i-n-fo rma t-i-o n-d ur-'i-n g-t he--s hutdown_modes_a nct ___________ _

during startup and shutdow11 operations. When independent verification of rod movement is required, the IRPIS will still be available to provide this information. The proposed changes are as follows:-

1. The IRPIS operability requirements during shutdown conditions have been removed. These requirements have been replaced by a requirement for a +/-2 step demand counter accuracy.

While the IRPIS was originally not intended to be used in the shutdown modes, many plants have a specification requiring IRPIS indication during these modes as an intended means of providing added confidence to the shutdown reactivity margin calculations. The reality of the situation is that, due to the transient effects, it is not possible to use the system accurately in these modes and therefore the use of the IRPIS may create more problems than it solves.

Concerning the issue of shutdown margin, the Technical Specifications require a minimum of 1.0% shutdown reactivity margin during shutdown (i.e., keff < 0.99), and station administrative procedures for determining the required shutdown boron concentrations in these modes already include conservative allowances for calculation arid measurement uncertainties. Further, the demand position indicators Page 3

e have been highly reliable, not only at Surry but throughout the industry, and are demonstrably the most accurate means of determining bank position. This high level of reliability, combined with the requirement that keff < 0.99 in shutdown conditions, provides adequate protection against inadvertent criticality in the shutdown

- -- -~.------modes-;- ---As-a----f:!e-Su-1-t,--the- -I Rf>JS_j_s_no_t___ n_e_ede_d ___t_Q_ __ g_y_a ran tee ___

  • _____ _

subcriticality and/or shutdown margin in these modes, and the use of the system could in fact be counterproductive to this end. Consistent with the Staff SER of Reference 1, we therefore propose the replacement of IRPIS tol~rance requirements with step demand counter tolerance requirements during shutdown.

2. The second feature is the incorporation of a 11 soak time, 11 of up to one hour in every twenty-four, be 1ow 50% power. The IRPIS channe 1 to 1erance may be re 1axed for up to one hour fo 11 owing rod motion.

During this hour, the step demand counters will serve as the primary indicators of rod position, with the IRPIS channels displaying general information regarding rod motion; i.e., they can still be used to verify that the rods are moving in or out on demand, but should not be relied on for precise position indication. We believe this proposa 1 is sound from a safety standpoint due to the fo 11 owing considerations:

a. The probability of a severe rod misalignment actually occurring is remote, because 1) the IRPIS tolerance remains at +/-24 steps for only an hour before reducing to +/-12 steps, 2) the step demand counter accuracy is excellent and 3) the IRPIS wil 1 be used to verify rod movement upon demand.

Page 4

b. Rods which are misaligned by +/-24 steps for up to an hour do not pose an unacceptable risk. We analyze the impact of s~atically misaligned or dropped rods on the core power distribution throughout core life for each reload core and demonstrate that the Condition II DNB limits are met for po~er levels up to and
  • ------'ine--luEl-ing~-hot--fu-1-1-*powel"-. ____Eur.ther-_,_~th_e~_p_roru:,sec;t_one-ho_!J_!'__ ~----

relaxation would only apply at low power (<50%). The probability of experi~ncing a limiting Condition II event (i.e., *uncontrolled rod withdrawal) during a one-hour interval is insignificant (on the order of 10-6).

c. The 11 one hour in twenty-four 11 feature provides an acceptab 1e upper limit on the frequency of 11 Thermal Soak 11 allowances. This feature ensures that the IRPIS is stil1 used a minimum of 96% of the tjme (23/24) with +/-12 step accuracy, even at low powers, while providing the plant operators with sufficient tolerance relief when it is needed. Sufficient rod position information is available both for a continuous one hour period or one consisting of several discrete intervals.

A detailed discussion and evaluation of the proposed T~chnical Specification changes, by each section, follows:

Technical Specification 3.12.C.1 - Inoperable Rod Definition The definition of an inoperable control rod has been revised to be consistent with the one hour "Thermal Soak" allowance. Previously, a rod showing misalignment in excess of +/-12 steps was to be declared inoperable; Page 5

now, the misalignment widens to +/-24 steps during the "Thermal Soak" period, and remains at +/-12 steps otherwise. In the shutdown modes, when the step demand counters are the primary source of rod position information, rods will be declared inoperable only when the IRPIS fails to show rod movement upon demand.

Technical Specification 3.12.E.l - Rod Position Indication Requirements The proposed Technical Specifications denote each operational mode and the rod position indication requirements for each mode. Those conditions are, respectively, power operation above 50% power; power ?Peration between 0% and 50% power;* and operation at shutdown conditions.

Above 50% power, there is no change to the current requirement that the IRPIS provide rod positions with an accuracy of +/-12 steps.

Between zero and 50% power, the +/-12 step IRPIS tolerance requirement remains, except for the one hour "Thermal Soak" period, at which time the tolerance doubles to +/-24 steps to allow for the thermal nonequilibrium 11 effects. During this Therma l Soak" a 11 owance, the step demand counters must indicate rod locations with an accuracy of +/-2 steps. The one hour period may b~ continuous or may consist of several discrete intervals, but the "Thermal S0ak 11 allowance may be employed for a total of no more than one hour in twenty-four.

In the shutdown conditions, responsibility for rod position indication shifts to the step demand counters, which are required to be accurate to

+/-2 steps. This change reflects the excellent performance of the counters Page 6

and the unimpressive IRPIS performance in these modes. The IRPIS remains a van ab 1e to verify the occurrence and direction of rod movement, however.

Technical Specification 3.12.E.2.a - IRPIS Inoperability The references to the excore detectors and the incore thermocouples have been deleted as available instruments for rod position indication when an IRPIS is out of service. Neither instrument i~ well suited for this task, and in practice neither has been used to determine rod locations.* In contrast, the incore detectors are very well suited for individual rod position indication, and remain the primary backup for this function.

Technical Specification 3.12 - Basis Section A discussion has been added to the Basis section which addresses the 11 Therma 1 Soak 11 a 11 owance for nonequil i bri um effects, noting its justification and limitations. The shift of responsibility for rod position indication to the step demand counters during the shutdown modes is also described.

Technical Specification Table 4.1 Surveillance Requirements Item #9 of this table specifies the surveillance requirements for the Analog Rod Position system. Previously, the plant staff was not required to perform channel checks when the plant was in cold shutdown; with the switch to the step demand counters for rod position indication in the Page 7

shutdown modes, the plant staff will be required to perform channel checks during power operation only.

Item #9 of this table specifies the surveillance requirements for the Analog Rod Position system. Previously, the plant staff was allowed to forego the requirements when the plant was in cold shutdown; with the switch to the step demand counters for rod position indication in the shutdown modes, the Table 4.1-1 exclusion has been re-written to be I

applicable in hot, intermediate and cold shutdow~.

To support the increased emphasis on the step demand counters, a new requirement for quarterly testing of the control bank counters has been added to Item #10 ( 11 Rod Position Bank Counters 11 ) of Table 4.1-1.

Page 8

III. 10 CFR 50.59 SAFETY EVALUATION The proposed changes have been determined not to pose an unreviewed safety question as defined in 10 CFR 50.59. The basis of this determination is as follows:

  • The proposed changes will increase neither the probability of occurrence of any of the UFSAR accidents nor their potential consequences. No new or unique accident precursors are introduced by reliance upon the demand counters for rod position indication.

Likewise the consequences of any accident will not increase. The risk of a potentially misaligned rod not being detected during the one-hour thermal soak time is inconsequential, especially in light of the fact that even during this hour the IRPIS will be available for verifiiation that the rods are moving on demand.

  • No new or unique accident precursors are introduced by the proposed changes since no physical plant changes are involved. Thus, the possibility of an accident of a type different from those already considered in the UFSAR is not created.
  • The margin of safety is not reduced. Since the results of the UFSAR accident analyses will continue to bound operation under the proposed changes, the existing safety limits remain inviolate and as such, there is no safety margin reduction.

Page 9

IV.

SUMMARY

AND CONCLUSIONS Virginia Power has develbped a proposed set of Technical Specifications changes for Surry Units I and 2 which are intended to improve operational flexibility by accounting for known inaccuracies in the Individual Rod Position Indication System.

The basic thrust of the proposed changes is to shift the emphasis from the IRPIS to the demand position indicators as the primary source of rod position indication during shutdown conditions and for one hour following rod motion at low power. The IRPIS still serves as a backup system and can always be used to verify the occurrence and direction of rod motion upon demand. There is no change to the current IRPIS tolerance requirement above 50% of rated thermal power.

Page 10

REFERENCES 11

1. NRC Safety Evaluation Report (SER), Westinghouse Analog Rod Position Indication for Shutdown Modes, 11 transmitted by l_etter from S. A.

Varga (NRC) to G. W. Giesler (Wisconsin Public Service Corporation),

March 24, 1983. See also the letter from Mr. Varga to R. E. Uhrig, 11 Requi rements for Ana 1og Position Instruments In Shutdown Modes-Turkey Point Units 3 and 4, 11 March 15, 1982.

11

2. NRC Safety Evaluation, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 51 to Facility Operating License No. DPR-66, Duquesne Light Company/Ohio Edison Company/

Pennsylvania Power Company, Beaver ~alley Power Station, Unit No. 1, Docket No. 50-334, 11 transmitted by Letter from P. S. Tam (USNRC) to J. J. Carey (Duquesne Light), June 14, 1982.

Page 11

ATTACHMENT 3 10 CFR 50.92 Evaluation

e 10 CFR 50.92 SIGNIFICANT HAZARDS CONSIDERATION Virginia Power has proposed changes to the Surry Power Station Technical Specifications which shift responsibility for tracking control rod position from the individual rod position indication system to the step demand counters. It has been determined that the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination may be stated as follows.

  • The proposed changes will increase neither the probability of occurrence of any of. the UFSAR accidents nor t~eir potential consequences. The probability of a misaligned rod rema1n1ng undetected is exceedingly small, because of ~he tolerance requirement and the historically excellent performance of the step demand counters, and because of the availability of the IRPIS to verify the occurrence and direction of rod movement upon demand. Further, since the proposed changes shift rod tracking responsibility to the step demand counters when the IRPIS is least reliable, rod control will be more precise and accident probability will if anything be reduced.

Neither wil 1 the potent i a 1 con sequences of any postulated accident increase. Peaking factors which occur as a consequence of severely misaligned or dropped rods are verified on a reload basis as not resulting in a violation of any safety limit; the. assumed misalignments easily bound any potential misalignment under the proposed Technical Specifications, so that these changes cannot result in an increase in an accident consequence.

  • No new or unique accident precursors are introduced by the proposed changes since no physical plant changes are involved. The procedural change is a move from less precise plant control to more precise control, without an attendant increase in procedural complexity or a change in hardware. Thus, the possibility of an accident of a type different from those already considered in the UFSAR is not created.
  • The margin of safety is not reduced. The results of the UFSAR accident analyses will continue to bound operation under the proposed changes, so that the existing safety limits remain inviolate.

Specifically, the peaking factor criteria during potential misaligned and dropped rod events will continue to be verified on a reload basis.

As such, there is no safety margin reduction.

l I

,.._ It ....

ATTACHMENT 4 Application Fee

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