ML20126D510

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Proposed TS SR 4.3.B.1 Re Coupling CRD to Control Rod
ML20126D510
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 12/18/1992
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20126D506 List:
References
NUDOCS 9212280023
Download: ML20126D510 (14)
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ATTACHMENT 1 to JPN 92 069 l

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PROPOSED TECHNICAL SPECIFICATION CHANGES i REVISION OF CONTROL ROD DRIVE  :

COUPLING INTEGRITY SURVEILLANCE REQUIREMENTS l

(JPTS 90 022) l t

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT-Docket No. 50 333 - i DPR 59 J 9212280023 921218-PDR- ADCWK 05000333 u _

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JAFNPP i

3.3 (cont'd)- 4.3 (cont'd) .

B. Control Rods B. Control Rods

1. Each control rod sha!! be coupled to its drive or completely 1. Demonstrate that each control rod drive does not go to the inserted ana the control rod directional control valves overtravel position:

disarmed electrically. This requirement does not apply in i the refuel condition when the reactor is vented. Two a. Each time a control rod is withdrawn to the " full out" i control rod drives may be removed as long as Specification position.  !

3.3.A.1 is met.

b. Prior to declaring a control rod OPERABLE, after work on a control rod or the CRD System that could affect coupling.
2. The control rod drive housing support system shall be in 2. The control rod drive housing support system shall be place during reactor power operation or when the reactor inspected after reassembly and the results of the inspection coolant system is pressurized abue atmospheric pressure recorded.

with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3.A.1 is met. i i

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,1 Amendment No.[1 ,

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-3.3 and 4.3 BASES (cont'd)' .

7-i the control cell geometry and local k . Therefore, an Also if damage within the contro! rod drive mechanism and i additional margin is included in the shutdown margin test to in particular, cracks in drive internal housings, cannot be j account for the fact that the rod used for the demonstration ruled out, then a generic problem affecting a number of i (the analytically strongest) is not necessarily the strongest drives cannot be ruled out. Circumferential cracks resulting 2

rod in the core. Studies have been made which compare from stress assisted intergranular corrosion have occurred -

experimental criticals with calculated criticals. These in the collet housing of drives at several BWRs. This type studies have shown that actual criticals can be predicted of cracking could occur in a number of drives and if the within a given tolerance band. For gadolinia cores the cracks propagated until severance of the collet housing additional margin required due to control cell material- occurred, scram could be prevented in the affected rods.

manufacturing tolerancesand calculational uncertaintieshas Limiting the period of operation with a potentially severed experimentally been determined to be 0.38% ak. When collet housing will assure that the reactor wi!I not be l this additional margin is demonstrated, it assures that the operated with a large number of rods with failed collet
reactivity control requirement is met. housings.
2. Reactivity Margin -Inoperable Control Rods B. Control Rods 4- ,

Specifkation 3.3.A.2 requires that a rod be taken out of 1. Couptmg venfication is performed to ensure the control rod service if it cannot be moved with drive pressure. If the rod is connected to the - Control Rod Drive (CRD). The is fully inserted, it is in a safe position of maximum Surveillance requires demonstrating a CRD does not go to -

contribution to shutdown reactivity. If it is in a non-fully the overtravel position when it is fully withdrawn. The inserted position, that position shall be consistent with the

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overtravel position feature provides a positive check on the shutdown reactivity limitation stated in Specification . coupling integrity since only an uncoupled CRD can reach j 3.3.A.1. This assures that the core can be shut down at all . the overtravel position. The verification is required to be l times with. the remaining control rods assuming the performed any time a control rod is withdrawn to the " full i strongest operable control rod does not insert.' out" (notch position 48) position or prior to declanng the I

i control rod to be OPERABLE after work on the control rod

- Inoperable bypassed rods will be limited within any group or CRD System that could affect coupling. This includes to not more than one control rod of a (5x5) twenty-five control rods inserted one notch and then retumed to the control rod array.'

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Amendment No. ,1 ,

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JAFNPP .

3.3 and 4.3 BASES (cont'd)

" full out' position during the performance of SR 4.3.A.2.a. 3. The Rod Worth Minimizer (RWM) restricts the order of This Frequency is acceptable, considering the low control rod withdrawal and insertion to be equivalent to the probability that a control rod will become uncoupled when Banked Position Withdrawal Sequence (BPWS). These it is not being moved, and operating experience related to sequences are established such that the drop of any uncoupling events. in-sequence control rod from the fully inserted position to the position of the control rod drive would not cause the .

reactor to sustain a power excursion resulting in a peak fuel

2. The control rod housing support restricts the outward enthalpy in excess of 280 cal /gm. An enthalpy of 280
  • movement of a control rod to less than 3 in. in the cal /gm is well below the level at which rapid fuel dispers2 extremely remote event of a housing failure. The amount could occur (i.e. 425 cal /gm.). Primary system damage in of reactivity which could be added by this small amount of this accident is not possible unless a significant amount of rod withdrawal, which is less than a normal single fuel is rapidly dispersed. Ref. Subsections 3.6.6, 7.7.4.3 '

withdrawal increment, will not contribute to any damage to and 14.6.1.2 of the FSAR, NEDE-24011 and NEDO-10527 '

the Primary Coolant System. The design basis is given in including supplements 1 and 2 to NEDO-10527.  ;

subsection 3.8.2 of the FSAR, and the safety evaluation is given in subsection 3.8.4. This support is not required if in performing the function described above, the RWM is not the Reactor Coolant System is at atmospheric pressure required to impose any restrictions at core power levels in since there would then be no driving force to rapidly eject excess of 10% of rated. Materialin the cited references a drive housing. Additionally, the support is not required if shows that it is impossible to reach 280 calc.ies per gram all control rods are fully inserted and if an adequate in the event of a control rod drop occurring at power shutdown margin with one control rod withdrawn has been greater than 10%, regardless of the rod pattem. This is demonstrated, since the reactor would remain subcritical true for all normal and abnormal patterns including those even in the event of complete ejection of the strongest which maximize the individual control rod worth.

control rod.

Amendment No. , ,

100 1

Attachment 11 to JPN 92 069 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REVISION OF CONTROL ROD DRIVE COUPLING INTEGRITY SURVEILLANCE REQUIREMENTS (JPTS 90-0221

1. DESCRIPTION OF THE PROPOSED CHANGES This application for an amendment to the James A. FitzPatrick Technical Specifications propose revised surveillance test requirements for determining control rod / Control Rod Drive (CRD) coupling integrity.

Minor changes in format, such as type font, margins or hyphenation, are not described in this submittal. These changes are typographicalin nature and do not affect the content of the Technical Specifications.

Paae 91. Soecification 4.3.B.1 Replace the current Surveillance Requirement with the following Surveillance Requirement:

"1. Demonstrate that each control rod drive does not go to the overtravel position:

- a. Each time a control rod is withdrawn to the " full out" position,

b. Prior to declaring a control rod OPERABLE, af ter work on a control rod or the CRD System that could affect coupling."

Panes 99 and 100. Bases Section 3.3 and 4.3 B.1 Replace the current contents of Section B.1:

" Control rod drop accidents as discussed in the FSAR can lead to significant core damage. If coupling integrity is maintained, the possibility of a rod drop accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive. Absenco 6-of such response to drive movement could indicate an uncoupled condition. Rod.

position indication is required for proper function of the Rod Worth Minimizer (RWM)."

.with the following:

" Coupling verification is performed to ensure the control rod is connected to the Control Rod Drive (CRD). The Surveillance requires demonstrating a CRD does not go to the overtravel position when it is fully withdrawn. The overtravel position feature providos a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position. The verification is required to

. be performed any time a control rod is withdrawn to the " full out" (notch nosition

48) position or prior to declaring the control rod to be OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control

Attachment 11 to JPN 93 069

. SAFETY EVALUATION Page 2 of 5 rods inserted one notch and then returned to the " full out" position during the performanco of SR 4.3.A.2.a. This Frequency is acceptable, considering the low probability that a control rod will becomo uncoupled when it is not being moved, and operating oxporlence related to uncoupling events."

J ll. PURPOSE OF THE PROPOSED CHANGES l

The CRD mechanisms (e.g., pistons, drives, collet components, etc.) are the means j by which the control rods are inserted and withdrawn to control core reactivity. The CRD can position the control rods in 6 inch increments by a slow controlled motion  ;

i (as in normal reactor startup and shutdown) or provide rapid insertion (scram) during an abnormal operating condition.

4 The proposed changes to the James A. FitzPatrick Tochaical Specifications revise the surveillance requirements for centrol rod /CRD coupling integrity based on guidance from the improved Standard Technical Specifications (STS) (Reference 1). Although the current surveillance requirement establishes control rod coupling integrity following a refueling outage or maintenanco,it does not provide for datormination of CRD coupling integrity overy time the rod is fully withdrawn. The purpose of this revision is to provide additional assuranco of control rod coupling by requiring a 4

coupling check overy time a control rod is fully withdrawn. Adoption of the improved STS achieves this purpose and clarifies the wording of the surveillance requirement.

1 Ill. SAFETY IMPLICATIONS OF THE PROPOSED CHANGES The current surveillance requireront confirms coupling integrity of a control rod during the initial withdrawal following a refueling outage, control rod or CRD maintenance. The only valid method of demonstrating control rod /CRD coupling integrity is when the CRD does not enter the overtravel position upon command. The revised surveillance requirement requires verification of control rod /CRD coupling integrity using the same methodology each time a rod is brought to the " full out" position. The removal of references to the nuclear instrumentation (i.e., noutron monitors) from the existing surveillance requirements for determining control rod /CRD coupling integrity does not raise any safety mplications since they will only indicate whether or not a control rod is stuck. Furthermore, there are situations in which a control rod could be completely withdrawn after refueling without causing a nuclear instrument responso.

Therefore, eliminating the use of nuclear instruments as a method of determining control rod /CRD coupling integrity will not reduce the margin of safety. The "uncouplod" events at other plants described in References 2 and 3 indicate that additional control rod surveillances provide an improved safety benefit in the ability to detect uncoupled control rods resulting from mechanical errors. The revised surveillance requirements will provide increased assurance of control rod /CRD coupling integrity through additional surveillanco.

Attachrnent il to JPN 92-069 SAFETY EVALUATION Page 3 of 5 The proposed chango differs from the improved STS by using the word "demonstrato" instead of the word " verify" in Specification 4.3.B.1 and the word

" demonstrating" instead of " verifying" in the second sentenco of Bases 3.3 and 4.3 B.1 on page 99. The Technical Specifications were revised in Reference 4. This revision interpreted the meaning of the word " verify" as "the associated surveillance activities have been satisfactorily performed within the specified timo interval" and the word " demonstrate" as " conduct a test to show." Replacing the words " verify and verifying" with the words "demonstrato and demonstrating," respectively, more accurately describes the surveillance requirements and providos consistency with the remainder of the Technical Specifications. This change will not reduce the effectiveness of the proposed requirements.

Other differences betwoon this submittal and the improved STO are editorial in nature to reflect plant specific technical specification style. The proposed Surveillance Requirements will not alter the conclusions of the plant accident analyses as documented in the FSAR c. the NRC's staff SER (References 5 and 6).

IV. ILVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordanco with the proposed Amendment would not involvo a significant hazards consideration as def;ned in 10 CFR 50.92, since it would not:

1. involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes increase the frequency but not the method 3 logy for testing control rod /CRD coupling integrity. The proposed changes involve no hardware changes, no changes to the operation of the CRD systems or method of coupling to the control rods. The proposed Technical Specification amendment does not alter the ability of the CRD systems in performing their intended function.

Thorofore, the changes have no effect on conditions that could affect the consequences of an accident. The increased testing for control rod /CRD coupling integrity will not result in an increased probability of an accident.

2. create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes willincrease testing frequency. They will not require modification of test processes or of any plant structures, systems, or components. The changes will not alter the CRD system function or operability.

The changes will not affect any conditions that could result in a new or different type of accident.

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Attachment 11 to JPN-93 069 l

. SAFETY EVALUATION l Page 4 of 5 '

3. involve a significant reduction in a margin of safety.

The proposed changes will not adversely affect the margin of safety. The changes provido an increased level of confidence in control rod /CRD coupling integrity and no reduction in any margins of safety.

V. IMP _LEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not adversely affect the ALARA or Firo Protection Programs at the FitzPatrick plant, nor will the changes affect the environment. The proposed changes to revise the control rod /CRD coupling integrity surveillanco requirements performed from the control room can havo no affect on any of those programs or the environment.

VI. CONCLUSION The changes, as proposed, do not constituto an unroviewed safety question as defined in 10 CFR 50.59. That is, they:

1. will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report:
2. will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report; and
3. will not reduce the margin of safety as defined in the basis for any technical specification.

The changes involve no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. NRC NUREG 1433 " Standard Technical Specifications in General Electric Boiling Water Reactors (BWR/4)," Revision 0, dated September 1992.
2. GE information letter, dated July 31,1974, Sll No. 52, Category 2, Supplement 2 "CRD Inner Filter Replacement."
3. GE information letter, dated March 17,1989, SIL No. 52, Category 1, Supplement 3, " Control Rod Uncoupling Rod Replacement."

4, NRC letter, D.E. LaBarge to J.C. Brons dated December 26,1989 (JAF-90 002) transmits Amendment 148.

, Attachmont il to JPN 93 069 SAFETY EVALUATION Page 5 of 5

5. James A. FitzPatrick Nuclear Power P. Updated Final Safety Analysis Report Section 3.5, " Control Rod Drive Mechanical Design," through Revision 5, dated January 1992.
6. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.

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l ATTACHMENT lli to JPN 92 009 l PROPOSED TECHNICAL SPECIFICATION CHANGES REVISION OF CONTROL ROD DRIVE COUPLING INTEGRITY SURVEILLANCE REQUIREMENTS MARKUP OF TECHNICAL SPECIFICATION PAGES i \

(JPTS 90 022) l 1

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 DPR 59

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' the control ceR geometry and local k, . Therefore, an '" '* *M addihonal mergin is irw* =*=ri in tie shutdown margin test , g to account for the fact the the rod used for the U demonstration . (the anefyticeNy strongest) is not Also if damage within the control rod drive mediastiam rw==tarily the strongest rod in the core. Studies have and in parbcular, cracks in drive ritemel houangs, been made which compere experimental cnbcais with cannot be ruled out, then a genenc problem apocting a ca h d=8ari cr e eman These ahne== have shown that number of dnwes cannot be ruled out. Circumferenhal areud cretscels can be pr=rer*=rt weihin a given tolerance cracks resulhng from stress asested irdergrander j band. For gedolinia cores the adcSlioned margin required corrosson have occurred in the collet houemg of drtwas et i due to control ces meterief rnamdar*=ing tolerances and several BWRs. This type of ciscJsg codd occur in a e mars d=*ianel uncertainhos ' has amperimentaNy been number of drives and if the cracks propagated ureil determined to be 0.38% Ak. Mihen viis arMaiarial rnergin severance of the coBot houesng occurred, scram could is demonstrated, it secures tiet the reactively control -

be presented in tie apocted rods. Umshng the penod of requwementis met. operation witt a potenheBy severed meet housing wig assure that tio reactor wiE not be operated utIh a large

2. .~_- ^' '.., % - W W Rods .g,,,,,,,,,

Speedicahon 33A2 requires that a rod be taken out of r sonna if it cannot be moved with drive pressure. If the B. Control Rods '

rod is fuNy inserted, it is in a safe poeshon of mammum contrirudinri to shutdoum reeceivity. It it is in a norWupy 1. Controlrptidrop as docussedin can inserted poeleon, that poolhon sher be consistent with lead 1(signlEcont damajgII. If the shutdown reacsevity limitasion stated in Specslication irj? d, Wisy-- _ S^pW a rod is 33A1. This amoures that the core can be shut dourn at ..;. T _1 The overtravel ^

aR times with the remaining control rods aneuming wie -

strongest operable coned rod does not insert.

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. o INSERT "A" Demonstrate that each control rod drive does not go to the overtravel position:

a. Each time a control rod is withdrawn to the " full out" position.
b. Prior to declaring a control rod OPERABLE, after work on a control rod or the CRD System that could affect coupling.

INSERT "B" Coupling verification is performed to ensure the control rod is connected to the Control Pod Drive (CRD). The Surveillance requires demonstrating _

e CRD does not go to the overtravel position when it is fully withdrawn.

The overtravel position feature provides a positive check on the coupling integrity since only on uncoupled CRD can reach the overtravel position.

The verification is required to be performed any time a control rod is withdrawn to the " full out" (notch position 48) position or prior to declaring the control rod to be OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the " full out" position during the performance of SR 4.3.A.2.a. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved, and operating experience related to uncoupling events.

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