Difference between revisions of "ML20071C388"

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Proposed Tech Spec Rev Renumbering Mechanical Snubbers on Table 3.6-lb for Clarity,Adding Sixteenth Snubber & Changing Emergency Preparedness Audit Frequencies
ML20071C388
Person / Time
Site: FitzPatrick Exelon icon.png
Issue date: 02/25/1983
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20071C376 List:
References
Download: ML20071C388 (9)


Text

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ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO SURVEILLANCE, AUDIT FREQUENCY AND ADMINISTRATIVE CHANGES l

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L POWER AUTHORITY OF THE STATE OF NEW YORK

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 l

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8303020021 830225 -

PDR ADOCK 05000333' PDR;

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TABLE 3.6-1b SAFETY RELATED MECHANICAL SNUBBERS g

$ HIGH RADIATION ESPECIALLY O' SNUBBER ACCESSIBLE OR ZONE DIFFICUCT

$ NO. LOCATION ELEVATION INACCESSIBLE DURING SHUTDOWN TO REMOVE e

,$ 10-7B-S3 Crescent Area RilR-45B 239' A NO NO 10-9A-S40 Crescent ?.rea MOV-65B 246' A YES NO l 10- 13A-S112 Reactor Building 315' A NO NO 12"-W20-302-13A .

10-15B-S124A Crescent Area 256' A 'M) NO 10-15B-S124B 16"-W20-302-15B 10-15B-S125A Crescent Area MOV-39B 256' A -

NO NO 10-15B-S125B 10-3B-S153 Crescent Area 232'- A NO NO

" 24"-W20-152-3B 10-35-S221A Reactor Building 336' A NO NO 4"-W20-302-35 ,

10-9A-S254 Crescent Area 262* A NO NO 16"-W20-302-9A 10-9B-S286 Crescent Area 253' A NO NO 16"-W20-302-9B

. 24 "-W20 - 15 2 - 3 B 10-15A-S307 Crescent Area MOV-39A 256' A NO NO 10-15A-S308 -

10-15B-S119A Crescent Area 256' A NO NO 16"-W20-302-15B ,

j 3.11 t at'd) JAFNPP 4.11 (cont'd) 1 B. Crescent Area Ventilation B. Crescent Area Ventilation l'

Crescent area ventilation and cooling Unit coolers serving ECCS

] equipment shall be operable on a components shall be checked .

continuous basis whenever specification for operability once/3 months

. 3.5.A, 3.5.B, and 3.5.C are required

> to be satisfied. 1. When it is determined that two unit coolers serving

1. From and after the date that ECCS components in the same more than one unit cooler compartment are made or 1 serving ECCS components in , found inoperable,. reactor the same compartment are made operation may continue for or found to be inoperable, 7 days unless one is made i

all ECCS components in that operable earlier.

compartment shall be considered

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Temperature indicator to be inoperable for purposes 2.

! of specification 3.5.A, 3.5.C, controllers shall be

! and 3.5.D. calibrated once/ operating cycle.

C. Battery Room Ventilation I 3. If 3.11.B.1 cannot be met, Battery room ventilation shall be the reactor shall be placed f

operable on a continuous basis in a cold coEidition within whenever specification 3.9.E is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

required to be satisfied.

C. Battery Room Ventilation

1. From and after the date that

! one of the battery room ventil- Battery room ventilation equip-ation systems is made or found ment shall be checked for l operability once/ week.

j to be inoperable, its associated l battery shall be considered to j be inoperable for purposes of 1. When it is determined that l specification 3.9.E. one battery room ventilation systesa is inoperable, the l remaining ventialtion system i shall be checked for operability and daily thereafter.

2. Temperature transmitters and differential $ressureswitches l sM11 M calhated once/ opera &g cycle.

Amendment No. g 239 f

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c. Th3 r0culto of cctiono teken to correct d3ficisncion occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at
least once per G months.
d. The performance of activities required by the Opera-tional Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

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e. The Facility Emergency Plan and implementing procedur?.s at least once per 12 months. l
f. The Facility Security Plan (including the Safegua'rds Contingency Plan) and implementing procedures at least once per 12 months.

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g. Any other area of facility. . operation considered appropriate by the SRC or the Senior Vice President--

Nuclear Generation.

e h.' The Facility Fire Protection Program and implementing procedures at least once per two years.

i. An independent fire protection and loss of prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel an outside~ fire protection firm. '
j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than
3 years.

l 6.5.2.9' AUTHORITY l

The SRC shall report to and advise the Senior Vice President Nuclear Generation on those areas of responsibility specified in Section 6.5.2.7 and 6.5.2.8.

6.5.2.10 RECORDS Records will be maintained in accordance with ANSI 18.7-1972. The following whall be prepared, approved and distributed as indicated below:

a. Minutes of each SRC meeting shall be prepared, approved and forwarded to the Senior Vice President-Nuclear-Generation l within 14 days after the date of the meeting.
b. Reports of review encompassed by Section 6.5.2.7 above shall be prepared, approved and forwarded to the Senior Vice President-Nuclear-Generation within 14 days following completion of the review.
c. Audit reports' encompassed by Section 6.5.2.8 above, shall be forwarded to the Senior Vice President-Nuclear-Generation and to the management positions responsible for the areas audited within 30 days after completion of the audit.

Amendment No. 5<f, fdf, g 252 a

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N VICE PRESIDENT

$ RESIDENT QUALITY ASSURANCE g MANAGER AND STAFF g J L

" PLANT OPERATING Z REVIEW COMMITTEE ~

]

.O SUPERINTENDENT .

OF POWER

< AND STAFF l

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I I MAIN- I &C OPERATIONS RADIOLOGICAL TECHNICAL SECURITY / SAFETY QA SUPT.

. TENANCE SUPT. SUPT. (SRO) & ENVIRON- SERVICES FIRE PROTECTION & STAFF SUPT. & & AND STAFF MENTAL SERVICES SUPT. &

  • SUPT. & STAFF STAFF STAFF SUPT. & STAFF STAFF i

REACTOR 4 w ANALYST SHIFT 1

  • TECHNICAL ADVISOR TRAINING SUPERINTENDENT OFFICE .

AND STAFF MANAGER

s SHIFT SUPERVISOR (SRO)

I SENIOR NUCLEAR FIGURE 6.2-1 OPERATOR (RO)

I POWER AUTHORITY OF THE STATE OF NEW YORK NUCLEAR CONTROL JAMES A. FITZPA't. RICK NUCLEAR POWER PLANT

! OPERATOR (RO) PLANT STAFF ORGANIZATION 1

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  • Responsibility for performance and

{ AUXILARY NUCLEAR monitoring of the fire protection

! OPERATORS OPERATORS program.

SRO - SENIOR REACTOR OPERATOR I RO - REACTOR OPERATOR i

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t ATTACHMENT II PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO .,

SURVEILLANCE, AUDIT FREQUENCY AND ADMINISTRATIVE CHANGES POWER. AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT l

DOCKET NO. 50-333 s

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S ction I - Deccription of th9 Chenann ,

Table 3.6-lb-(page 156r, " Safety Related Mechanical Snubbers")

has been revised to incorporate a minor renumbering of the snubbers and the addition of a sixteenth snubber.

Figure 6.2-1, " Plant Staff Organization" on page 260 is updated to show that the Security and Safety Superintendent has responsibility for performance and monitoring of the fire protection program, in accordance with our previous commitment.

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Section 4.11.B (page 239, " Crescent Area Ventilation") is revised to require that unit coolers serving ECCS components are checked for operability once per three months, rather.than during

! the surveillance testing of the associated pumps, as currently

! required.

In Section 6.5.2.8(e) on page 252a, the audit frequency for emergency preparedness and safeguards contingency plans is revised to agree with the requirements of parts 73.40(d) and 50.54(t) of Title 10 of the Code of Federal Regulations.

Section II - Purpose of the Changes The changes to Table 3.6-lb will update this list of safety related snubbers to include a new, sixteenth unit. In addition, one of the snubber numbers has been changed for clarity. (Please note that this same page was changed in a previous amendment application dated July 13, 1981. As of this date, this application amendment i

has not been issued by the Commission. The page included in Attachment II of this application supercedes the previously submitted page.)

One of the two changes to Figure 6.2-1 incorpcrates an existing NRC commitment into the FitzPatrick Technical Specifications regarding the Security and Safety Superintendent's job function.

The second changes the " Training Coordinator" block to read

" Training Superintendent and Staff" to agree with current Authority organization and titles.

The current Section 4.11.B (Crescent Area Ventilation) requires excessive testing, is inconsistent with ALARA and does not improve reliability. The proposed test frequency (once every 3 months) corresponds to the frequency stated in the Inservice Inspection Valve Test Program for unit cooler temperature control valves. The present specification results in several tests of unit coolers each month.. This is because current specifications require unit cooler tests with each LPCI, Core Spray and HPCI operability and flow rate test. Personnel performing this testing are therefore required to l

spend additional time in the radiation areas in which these unit I coolers are located. Furthermore, there are heat loads in the i f crescent areas which require that these unit coolers function to l

maintain normal temperatures. These heat loads include ambient

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1 (wcra woathar) courcso, continuously heated stson supply linsa to HPCI and RCIC,.and non-emergency operation of ECCS components (e.g.

for surveillance testing, shutdown cooling or suppression pool cooling and level maintenance.) Thus, crescent area temperature provides a continuous indication of unit coolers operability. A

quarterly surveillance test adequately ensures reliability.

i In Section 6.5.2.8(e) (page 252a), the audit frequencies for emergency preparedness and safeguards contingency plans 1 (References f and g) are revised to agree with Title 10 of the Code of Federal Regulations, parts 50.54(t) 73.40(d), respectively. We were notified of the need for this change via Reference f and g.

l This is clarifying, and made at the request of the Commission.

! Section III - Impact of the Changes Figure 6.2-1 and Table 3.6-lb will be updated to reflect changes already in effect at FitzPatrick and therefore will not impact plant operation.

The reduced surveillance testing required by Section 4.11.B will reduce personnel radiation exposure while maintaining a high level of reliability and operability for these unit coolers.

These changes will not have any impact on the environment. The changes as proposed will not impact the Fira Protection Program at the James A. FitzPatrick Nuclear-Power Plant.

Section IV - Implementation of the Change I

Reduced personnel radiation exposure will result from reduced surveillance test requirements.

Section V - Conclusion The incorporation of these changes: a) 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; b) will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification; and d) does not constitute an unreviewed safety question.

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'Section VI - R7fercncon (a) Jcoto A. FitzPatrick Nuclear Pow 3r Plant Final Shfoty Analycio Report (FSAR).

(b) James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).

(c) J.P. Bayne (PASNY) to T.A. Ippolito (USNRC) dated July 13, 1981 regarding Proposed Technical Specification Changes Related to Snubber Surveillene (JPN-81-51).

(d) C.A. McNeill, Jr. (PASNY) to R.C. Haynes (USNRC) dated May 28, 1982 regarding I.E. Bulletin No. 81-01, (JAFP-82-0577).

(e) I.E. Bulletin No. 81-01, Surveillance of Mechanical Snubbers" dated January 27, 1981.

(f) Generic Letter No. 82-17, D.G. Eisenhut (NRC) to all Licensees dated October 1, 1982. . j (g) Generic Letter No. 82-23, D.G. Eisenhut (NRC) to all Licensees dated October 30, 1982.

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