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Proposed Tech Specs Changing Table 3.7-1 Re Primary Containment Isolation Valves
	ML20248A321
Person / Time
Site:	FitzPatrick
Issue date:	05/31/1989
From:	; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	;
Shared Package
ML20248A312	List: JPN-89-033, Forwards Application for Amend to License DPR-59,revising Tech Spec Table 3.7-1, Process Pipeline Penetrating Primary Containment, to Accurately Reflect as-built Condition of Plant & Promote Consistency W/Nrc STS; ML20248A314; ML20248A321;
References
	JPTS-84-005, JPTS-84-5, NUDOCS 8906080102
	Download: ML20248A321 (67)
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ATTACHMENTS TO JPN-89-033 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS REGARDING TABLE 3.7-1 PRIMARY CONTAINMENT ISOL.ATION VALVES (JPTS-84-005)

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 f[k60g0[j[83 3 P

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ISOLATION SGNAL CODES for Tcble 3.7-1 1:

Signal' Description -

A* Reactor vessel low water level - This is the highest of the three low water level signals. A scram occurs at thislevel.

B* Reactor vessel low-low-low water level - This is the lowest of the three low water level signals.

C* High radiation-main steamline D* Line break - main steam (high steam flow) ,

E* Line break - main steam (tunnel high temperature)

F* High drywell pressure G Reactor vessel low-low-low water level or high drywell pressure H* Containment high radiation J* Line break in Reactor Water Clean-up System - high area temperature K* Line break in RCIC System steam line to turbine (high steam line area temperature, high steam flow, low steam line pressure, or high turbine exhaust diaphragm pressure)

L* Line break in HPCI System steam line to turbine (hlgh steam line space temperature, high steam flow, low steam line pressure, or high turbine exhaust diaphragm pressure)

M* Low condenser vacuum in rur mode only P* Low main steam line pressure at inlet to main turbine (RUN mode only)

R* . Remote manual switch from control room (automatic Group A & B isolaton valves are capable of remote manual operation from the control room)

S Low drywell pressure T Low reactor pressure permissive to open Core Spray and RHR-LPCI valves U* High reactor vessel pressure or high drywell pressure !

V High temperature at outlet of Reactor Water Clean-up system non-regenerative heat exchanger X* Reactor vessel low water level or high drywell pressure when both shutdown cooling suction valves (10MOV-17 and 10MOV-18) are not fully closed and reactor vessel pressure is below 75 psig Y Standby Liquid Control System actuated Z* Reactor building ventilation exhaust high radiation

These are the isolation functions of the Primary Containment and Reactor Vessel isolation -

Control System; other functions are given for information only. Setpoints for isolation signals are located in Tables 3.2-1 and 3.2-2.

Amendment No.

207 l

__ O

. JAFNPP Notes For Table 3.7-1

PRIMARY CONTAINMENT ISOLATION VALVES" (Sh.1 of 2)

1. Main steam isolation valves require that both solenoid pilots be de-energized to close valves. Accumulator air pressure plus spring force act together to close valves when both '

pilots are de-energized. Voltage failure et only one pilot does not cause valve closure.

2. Primary containment spray and pressure suppression chamber (torus) cooling valves have interlocks that allow them to be manually reopened after automatic closure. This provision permits containment spray, for high drywell pressure conditions, and/or pressure suppression chamber water cooling. When automatic signals are not present, these valves may be opened for test or operating convenience.

3. Testable check valves are designed for remote opening with zero differential pressure across the valve seat. The valves close on reverse flow even through the test switches may be positioned for open. The valves open when pump pressure exceeds reactor pressure even though test switch may be positioned for close.

4. Control rod hydraulic lines are isolated by solenoid operated valves (normal insert and withdraw), air operated valves (scram inlet and outlet) and check valves (control rod drive cooling nater). The !ines that extend outside the primary containment are small and terminate in a system designed to prevent outleakage (i.e., the scram discharge volume and associated instrument volumes). Solenoid operated valves are normally closed, but open on normal rod movement (insert or withdraw). Air operated valves are normally closed, but open for control rod scram. Check valves are normally open, but closed during control rod scram and insert. All valve numbers shown are prefixed with the system and control rod drive hydraulic control unit designation. For example, the complete designation for control rod 26-27 scram inlet valve is 03-26-27AOV-126.

5. The standard minimum closing rate for automatic motor operated isolation valves is based on a nominal line size of 12 inch. Using the standard closing rate, a 12 inch line is isolated in 60 seconds.

6. Coincident signals "G" and "T" open valves. Special interlocks permit testing these valves by manual switch except when automatic signals are present.

7. Normal status position of valve (open or closed) is the position during normal power operation of the reactor and is provided in this table for information only (see " Normal Status" column).

8. Signal "A" or "F" causes automatic withdrawal of TIP probe. When the probe is withdrawn, the ball valve automatically closes by mechanical action.

9. Reactor building ventilation exhaust high radiation signal "Z" is generated by two trip units.

This requires one unit at high trip or both units at down scale (instrument failure) trip, in ;

order to initiate isolation.

10. Leak testing shall be accomplished in accordance with Section 4.7.A.2.d.

l l

Amendment No.J4(

208 I

_ - ____-___-_ A

r JAFNPP c

Notes For Table 3.7-1

PRIMARY CONTAINMENT ISOt.ATION VALVES" .

(Sh.2 of 2) -

11. Valve 20AOV 95 opens during pump out of the drywell equipment sump. Automatic ~l Isolation signals A and F override en open signal that might be present for sump pump out.

12. ' Radiation monitors used for sampling iodine, particulate, and gaseous are as follows:

Radiation Monitors Sample 17 RM-101 A lodine 17-RM-101B 17-RM-102A Particulate 17-RM-102B 17-RM-103A Gaseous 17-RM-103B

13. Isolation signals A, F, and Z may be manually overridden by keylock switch on the '

Monitoring and Analysis Panel (MAP) located in the relay room.

14. Traversing in-core Probe (TIP) penetrations are isolated by a guide tube and valve assembly which includes a solenoid operated ball valve and an explosive shear valve designed to sever and seal the TIP tubing and TIP detector helix. The guide tube and valve assemblies are designated 07NM-104A, B, C, and D for penetrations 35A, B, C, and D, respectively.

15. The explosive shear valves are not normally actuated and require replacement parts and maintenance activity in order to open the valves following actuation.

Amendment No.gg i

-m____ - . _ _ _

I-

.g .

ATTACHMENT 11

'TO JPN-89-033 SAFETY EVALUATION FOR PROPOSED CHANGES RELATED TO TECHNICAL SPECIFICATIONS TABLE 3.7-1 PRIMARY CONTAINMENT ISOLATION VALVES (JPTS-84-005) i~

NEWYORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

1 Attachment 11 to JPN-89-033 I SAFETY EVALUATION j

. Page 1 of 47 s Section i DESCRIPTION OF PROPOSED CHANGES l

This application for amendment to the FitzPatrick Technical Specifications proposes to clarify and update Table 3.7-1 (" Process Pipeline Penetrating Primary Containment", on pages 198 l through 209). The notes associated with this table are also revised. I To simplify the disc;ssion of the proposed changes to Table 3.7-1 and associated Notes, the changes are grouped into four categories. Some penetrations may be listed under more than one category. The categories are:

Category A- General Changes General changes clarify wordirig, improve nomenclature, and eliminate typographical errors currently existing in the Technical Specifications. These changes include:

. a better description of the penetration's function and valve designation; e changes that are compatible with the existing requirements in the NRC's Standard Technical Specifications; e elimination of unnecessary abbreviations; and

. elimination of ambiguities to reduce the potential for misinterpretation.

Category B - Addition of Valves Changes in this category include corrections to reflect the current condition of the plant.

The addition of valves to the table will complete the list of containment isolation valves.

This category is subdivided into two subcategories:

B1) Originalplant design (as-built)

Certain isolation valves were inadvertently omitted from the initial issuance of the Final Safety Analysis Report or the Technical Specifications. These valves have been part of the containment isolation system since the issuance of the Operating License.

B2) Plant modifications part of the NRC's " Post-TM! Requirements" program This category includes new valves added as a result of post-TMI modifications installed in accordance with NUREG-0578 and NUREG-0737.

Category C - Addition of isolation Signals This category is subdivided into two subcategories:

C1) Originalplant design (as-built)

This subcategory includes the addition of isolation signals to containment isolation valves as originally designed at the FitzPatrick plant. Certain isolation signals were omitted from the initial issuance of the FSAR or Technical Specifications. These signals have been part of the containment isolation system since the issuance of the Operating Ucense.

4

_____._______w

Att: chment il to JPN-89-033

(, SAFETY EVALUATION Page 2 of 47 l

C2) Post-TMImodifications 1

( This subcategory involves changes to reflect the implementation of post-TMI

! modifications.

Category D - Valve Closure Time Changes These changes reflect valve time adjustments performed to meet the requirements of the environmental qualification program.

A detailed description of these changes follows:

General Changes to Table 3.7 Category A

[a] The title " Process Pipeline Penetrating Primary Containment" is replaced by a new title,

Primary Containment Isolation Valves."

[b] Table entries are arranged by containment penetration number. In addition, six pages (Pages 203a through 206f) were added to the table to accommodate all the changes.

[c] The following columns are deleted: Valve Type; Power to Open; Group; Location Ref, to Drywell; and Power to Close.

[d] The column titled " Valve Type" is replaced with " Valve Number."

[e] Table headings for "Drywell Penetration," "Une Isolated," and " Remarks and Exceptions" are renamed " Containment Penetration," " Penetration Function," and

" Remarks," respectively.

[f] The abbreviation " secs" is replaced with the word seconds.

[g] The word "(Sec)", which is the abbreviation for seconds, is added to the Close Time column heading.

[h] isolation signal RM is changed to R.

[i] Power source abbreviations A-c and d-c are changed to AC and DC, respectively.

[j] Entries marked with dashes (" ") are replaced with *N/A".

[k] Containment Penetrations 30A and 50C are replaced with 30a and 50c, respectively.

[1] Containment Penetration 35B " Traversing in-core Probe Purge" is corrected to read 35E.

[m] Containment Penetration 36, " Control Rod Hydraulic Return," is no longer in service. This penetration has been cut and capped.

.t:,

Attachment il to JPN-89-033 SAFETY EVALUATION

~'

Page 3 of 47 9

]

Changes to Remark's Column of Table 3.7 Category A

[n]- Containment Penetration 10: Delete the note in the Remarks column.

[o] Containment Penetration 11: Delete the note in the Remarks column.

[p] Containment Penetration 23: Add the words "To drywell cooler assembly A and equipment sump cooler."

[q) Containment Penetration 24: Add the words "To drywell cooler assembly B."

~[r] Containment Penetrations 31Ad & 31Bd: Add the following remark as shown below:

Containment Penetration Remarks 31Ad From elev. 276' to Radiation Monitors. Note 12.

31Bd From elev. 296' to Radiation Monitors. Note 12.

.[s] Containment Penetrations 37A,378,37C, & 37D: Revise remark on control rod drive inlet valves as follows:

Typical of 107 Control Rod Hydraulic Control Units. Valves listed isolate the CRD insert and scram inlet line. Note 4.

[t] Containment Penetrations 38A,38B,38C, & 38D: Revise remark on control rod l . drive outlet valves as follows:

L Typical of 157 Control Rod Hydraulic Control Units. Valves listed isolate the CRD withdraw and scram outlet line. Note 4.

I L -- _ _ ____ _ _ _o

Attachment 11 to JPN-89-033 SAFETY EVALUATION A

, Page 4 of 47 1

s. . , ..

[u] Containment Penetrations 52a and 55b: Add the following remark as shown below:

g Containment Penetration Remarks 52a From Radiation Monitors to elev. 282'. Note 12.

55b From Radiation Monitors to elev. 296'. Note 12.

l

. . . . i

[v] Containment Penetrations 62 - 68: Add remarks to the penetrations as follows:

Containment Penetration Remarks 62 From drywell cooler assembly B.

63 To recirculation pump A and motor coolers.

64- From recirculation pump A and motor coolers.

65 From drywell equipment sump cooler.

66 From drywell cooler assembly A.

67 To recirculation pump B and motor coolers.

68 From recirculation pump B and motor coolers.

, . Attachment ll to JPN-89-033 H SAFETY EVALUATION

... Pige 5 of 47 3

[w] Containment Penetration 210A: Add remarks to valves as follows:

Valves Remarks

- 10MOV-34A Throttle valve for flow test and suppression pool cooling.' Note 2. .

- 13MOV-27 Pump minimum flow.

14MOV-5A Pump minimum flow.

14MOV-26A Throttle valve for flow test.

. . . . e

-[x] Containment Penetration 210B: Add remarks to valves as follows:

Valves Rema'r ks 10MOV-348 Throttle valve for flow test and suppression pool cooling. Note 2.

23MOV.25 Pump minimum flow.

14MOV 58 Pump minimum flow.

14MOV-26B Throttle valve for flow test.

Changes to Notes for Table 3.7 Category A

[y] The Notes for Table 3.7-1 are rearranged to start with Note 1 rather than with the isolation signal codes. The definition for isolation signal codes are now shown at the end of Table 3.7-1. In addition, setpoints for isolation signals will be referenced with the asterisk (*) footnote at the bottom of the signal code table which reads "Setpoints can be found in Technical Specification Tables 3.21 and 3.2-2."

[z] The parenthetical word "(torus)" is added to the first sentence of Note 2 (page 208).-

Attachment 11 to JPN-89-033 SAFETY EVALUATION

- Page 6 of 47 a

[aa) Note 4 (page 208) is completely revised and expanded to read:

Control rod hydraulic lines are isolated by solenoid operated valves (normal insert and withdraw), air operated valves (scram inlet and outlet) and check valves (control rod drive cooling water). The lines that extend outside the primary containment are small and terminate in a system designed to prevent outleakage (i.e., the scram discharge volume and associated instrument volumes). Solenoid operated valves are normally closed, but open on normal rod movement (insert or withdraw).

Air operated valves are normally closed, but open for control rod scram. Check valves are normally open, but closed during control rod scram and insert. All valve numbers shown are prefixed with the system and control rod drive hydraulic control unit designation. For example, the complete designation for control rod 26-27 scram inlet valve is 03-26-27AOV-126.

[ab] Notes 5,6,8,9 (which is blank),12, and 13 are deleted.

[ac) Note 7 is renumbered Note 5 and the abbreviations "in." and "sec" are replaced with " inch" and " seconds," respectively.

[ad] Note 10 is renumbered Note 6.

[ae] Note 11 is renumbered Note 7 and rephrased to read as follows:

7. Normal status position of valve (open or closed) is the position during normal power operation of the reactor and is provided in this table for information only (see " Normal Status" column).

[af] Note 14 is renumbered Note 8 and the word " ball" is added to the last sentence in the note.

[ag] Note 15 is renumbered Note 9, and the word " required" is changed to " requires" in

< the second sentence.

I

[ah) Note 16 is renumbered Note 10.

[ai] Note 17 is renumbered Note 11. In addition, the phrase "The valve opens .. " is replaced with the phrase " Valve 20AOV-95 opens ..."

[a)) The description for signal "A" in the isolation signal code definitions at the end of the table is rearranged to read as follows:

A* f.eactor vessel low water level - This is the highest of the three low water level signals. A scram occurs at this level.

1

i

Attachment 11 t)JPN-89-033 SAFETY EVALUATION Page 7 of 47 5

[ak] Isolation signal "D" is revised to read as follows:

D* Line break - main steam (high steam flow)

[al] Isolation signal "E" is revised to read as follows:

E* Une break - main steam (tunnel high temperature)

[am] Isolation signal "G" is revised to read as follows:

G Reactor vessel low-low-low water leve! or high drywell pressure

[an] Containment isolation signal code "H" is added to the list of isolation signal codes and will read as followe:.

H* Containment high radiation

[ao] The word " space" is replaced with the word " area" in the second sentence for containment isolstion signal "J".

[ap] The word " diaphragm" is added after exhaust in both for containment isolation signals J and K.

[aq) Containment isolation signals M and X are added to the list of isolation signal codes at the end of the table and they will read as follows:

M* Low condenser vacuum in run mode only X* Reactor vessel low water level or high drywell pressure when both shutdown cooling suction valves (10MOV-17 and 10MOV-18) are not fully closed and reacte vessel pressure is below 75 psig-close shutdown cooling injection valves (10MOV-25A and 10MOV-25B)

[ar] Isolation signal "RM" is changed to "R" and rephrased to read as follows:

R* Remote manual switch from control room (automatic Group A & B isolation valves are capable of remote manual operation from the control room) i

_ - __-_-___D

' Attachment 11 t)JPN-89-033 1 SAFETY EVALUATION

. -* Page 8 of 47-

[as] An asterisk (*) is added to isolation signal "U" and rephrased to read as follows: -

m.

i U* High reactor vessel pressure or high drywell pressure

- close shutdown cooling suction and reactor head spray valves

-[at] - A new Note 12 is added to page 209 and reads as follows:

12. Radiation monitors used for sampling lodine, particulate,

and gaseous are as follows

Radiation Monitors Sample 17-RM-101A lodine 17-RM-101B 17-RM-102A Particulate 17-RM 1028 17 RM-103A Gaseous 17 RM-103B

[au) A new Note 13 is added to page 209 and reads as follows:

13. Isolation signals A, F, and Z may be manually overridden by keylock switch on the Monitoring and Analysis Panel (MAP) located in the relay room.

[av) A new Note 14 is added to page 209 and reads as follows:

14. Traversing in-core Probe (TIP) penetrations are isolated by a guide tube and valve assembly which includes a solenoid operated ball valve and an explosive shear valve designed to sever and seal the TIP tubing and TIP detector helix. The guide tube and valve assemblies are designated 07NM-104A, B, C, and D for penetrations 35A,358,35C, and 35D, respectively.

c Attachment 11 to JPN-89-033 SAFETY EVALUATION

'v Page 9 of 47 7, .

lq [aw] A new Note 15 is added to page 209 and reads as follows:

15. The shear valves are explosively actuated and require replacement parts and maintenance activity in order to open the valves following actuation.

b Addition of Valves To Table 3.7 Category B Category B1 - OriginalPlant Design

-[ax] - Containment Penetration 11: add motor-operated valve 23MOV-60 with isolation signal "L,R," closure time "10," and normal status "Open."

[ay) . Containment Penetrations 25 & 71: add the following valves:

. - 27AOV-131 A with isolation signals "A, F, H, R, Z," closure time "5," and normal status " Closed."

. 27 CAD-68 with isolation signal " Reverse Flow" closure time *N/A," and normal status " Closed."

. 27AOV-131B with isolation signals "A, F, H, R, Z," closure time "5," and normal status " Closed."

. 27 CAD-69 with isolation signals " Reverse Flow" closure time "N/A," and normal status " Closed."

[azl Containment Penetrations 37A,378,37C,37D: The following valves are added:

. scram valve AOV-126 with isolation. signal "N/A," closure time *N/A," normal status " Closed," and remarks as shown in Category A.

. cooling water check valve CRD-138 with isolation signal " Reverse Flow,"

closure time "N/A," normal status "Open," and remarks as shown in Category A.

[ba] Containment Penetrations 38A,38B,38C,38D: add scram valve AOV-127 with isolation signal "N/A," closure time "N/A," normal status " Closed," and remarks as shown in Category A.

Attachment il to JPN-89-033 SAFETY EVALUATIOl1 e Page 10 of 47

[bb] Containment Penetration 39A: add containment spray air test isolation valve 10RHR-52A with isolation signal "N/A," closure time *N/A," and normal status

" Closed."

[bc] Containment Penetration 39B: add containment spray air test isolation valve 10RHR-528 and with isolation signal "N/A," closure time "N/A," and normal status

" Closed."

[bd] Containment Penetration 202A is deleted from Table 3.7-1.

[be] Containment Penetration 202B is revised to indicate that isolation valves 27AOV-101A and 27AOV-101B, previously on penetration 202A, with isolation signal

R,"

closure time "N/A," and normal status " Closed" are associated with 2028 only.

The description of " Valve" is replaced with "VB" (Vacuum Breaker) in the Remarks Column 4

[bf] Containment Penetration 211 A: add valve 10MOV-39A with isolation signals "G,R,"

closure time "N/A," normal status " Closed," and remarks " Note 2."

[bg] Containment Penetration 211B: add valve 10MOV 39B with isolation signals "G,R,"

closure time "N/A," normal status " Closed," and remarks " Note 2."

[bh] Containment Penetration 210A: add the following valves;

. 10MOV-16A with isolation signal "R," closure time "N/A," normal status

" Closed," and remarks of " Pump minimum flow";

. 10MOV-21 A with isolation signal "G,R," closure time *N/A," normal status

" Closed," and remarks " Heat exchanger drain";

. 10MOV-167A with isolation signal *R," closure time "N/A," normal status

" Closed," and remarks " Heat exchanger vent."

[bi] Containment Penetration 210B: add the following isolation valves;

. 10MOV-16B with isolation signal "R," closure time "N/A," normal status

" Closed," and remarks " Pump minimum flow";

. 10MOV-21B with isolation signal "G,R," closure time "N/A," normal status

" Closed," and remarks " Heat exchanger drain"; and

. 10MOV-167B with isolation signal "R," closure time "N/A," normal status

" Closed," and remarks " Heat exchanger vent."

Attachment 11 to JPN-89-033 SAFETY EVALUATION 9 Page 11 of 47 p

[b)) Containment Penetration 212: add valve 13RCIC-05 with isolation signal " Reverse -

Flow," close time "N/A," and normal status " Closed."

- [bk] Containment Penetration 214: add valve 23HPI-65 with isolation signal " Reverse Flow," close time "N/A," and normal status " Closed?

. [bl] L Containment Penetration 220: add the following valves:

. 27AOV-132A with isolation signals "A, F, H, R, ~Z," closure time "5," and normal status " Closed."

. 27 CAD-67 with ' isolation signal " Reverse Flow," closure time of "N/A," and normal status " Closed."

. 27AOV-132B with isolation signals "A, F, H, R, Z," closure time "5," and normal status " Closed."

. 27 CAD-70 with isolation signal " Reverse Flow," closure time "N/A," and normal status " Closed."

[bm] Containment Penetration 221: add the following valve:

. 13RCIC-07_ with isolation signal " Reverse Flow," closure time "N/A," and normal status " Closed."

' [bn] Containment Penetration 225A: add the following valves:

. 10MOV-13A and 13MOV-13C with isolation signal "R," close time "N/A," and normal status "Open."

[bo] Containment Penetration 225B: add the following valves:

. 10MOV-13B and 13MOV-13D with isolation signal "R," close time "N/A," and normal status "Open."

[bp] Containment Penetration 228: add penetration number 228 and valve 33CND-102 with isolation signal *N/A," closure time "N/A," and normal status " Closed."

t L_ _ . _ _ _ .

Attachment 11 to JPN-89-033 SAFETY EVALUATION Page 12 of 47 Category B2 - Post-TMI Modifications TMI-Item II.E.4.1 " Dedicated Hydrogen Penetrations"

[bq) Containment Penetration 26A & 26B: add valve 27MOV 122 with isolation signals "A, F, H, R, Z," closure time "5," and normal status " Closed." r

[br] Containment Penetration 205: add isolation valve 27MOV-123 with Isolation signals "A, F, H, R, Z," closure time "5," and normal status " Closed."

l TMI-Item 112.4.2 " Containment Isolation Dependability"

[bs)' Containment Penetration 23: add air operated valve 15AOV-130A with isolation signal "R," closure time "N/A," normal status "Open," and remarks as shown in Category A. .

[bt) Containment Penetration 24: add air operated valve 15AOV-130B with isolation signal *R," closure time "N/A," normal status "Open," and remarks as shown in Category A.

[bu] Containment Penetration 62: add air operated valve 15AOV-131B with isolation signal "R," closure time "N/A," normal status "Open," and remarks as shown in Category A.

i

[bv] Containment Penetration 63: add air operated valve 15AOV-132A with isolation ,

signal "R," closure time "N/A," normal status "Open," and remarks as shown in '

Category A.

[bw] Containment Penetration 64: add air operated valve 15AOV-133A with isolation signal "R," closure time *N/A," normal status "Open," and remarks as shown in Category A.

[bx] Containment Penetration 65: add penetration number 65 with; air operated valve ;

15AOV-134A with isolation signal *R," closure time "N/A," normal status "Open";

manual valve 15RBC-26A with signal "N/A," close time of and remarks as shown in Category A. I 1

)

Attachment il to JPN-89-033 SAFETY EVALUATION ;

l- .

Page 13 of 47 !

l l

1

[by] Containment Penetration 66: add air operated valve 15AOV-131A with isolation signal *R," closure time "N/A," normal status "Open," and remarks as shown in l Category A. I

(

[bz] Containment Penetration 67: add air operated valve 15AOV-1328 with isolation i

signal "R," closure time "N/A," normal status "Open," and remarks as shr>wn in l Category A.

[ca] Containment Penetration 68: add air operated valve 15AOV-133B with isolation signal "R," closure time "N/A," normal status "Open," and remarks as shown in Category A.

TMI-Item II.F.1.6 " Containment Hydrogen Monitor"

[cb] Containment Penetration 26A: add the following valves:

. 27SOV-119F2 with isolation signals "A, F R, Z," closure time "N/A," normal status " Closed," and remarks *From elev. 250'-3" to 27-PCA-1018. Note 13."

. 27SOV-119F1 with isolation signals "A, F, R, Z," closure time *N/A," normal status " Closed," and remarks "From elev. 250'-3" to 27-PCA-101B. Note 13."

. 27SOV-120E2 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From elev. 310 ~-6" to 27-PCA-101 A. Note 13."

. 27SOV-120E1 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From elev. 310'-6" to 27-PCA-101 A. Note 13."

. 27SOV-122E2 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "From elev. 343' to 27-PCA-101 A. Note 13."

. 27SOV-122E1 with isolation signals "A, F, R, Z," closure time *N/A," normal status " Closed," and remarks "From elev. 343' to 27-PCA-101 A. Note 13."

Attachment il to JPN-89-033 SAFETY EVALUATION Page 14 of 47

[cc] Containment Penetration 58b: add the following valves:

. 27SOV-122F2 with isolation signals "A, F, R, Z," closure time *N/A," normal status " Closed," and remarks *From elev. 343'-3" to 27-PCA-101B. Note 13."

. 27SOV-122F1 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "From elev. 343'-3" to 27-PCA-1018. Note 13."

[cd] Containment Penetration 58c: add the following valves:

. 27SOV-120F2 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From elev. 310'-6" to 27-PCA-101B. Note 13."

. 27SOV-120F1 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From elev. 310'-6" to 27-PCA-101 B. Note 13."

[ce] Containment Penetration 58d: add the following valves:

. 27SOV-123F2 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "From elev. 276'-6" to 27-PCA-101B. Note 13."

. 27SOV-123F1 with isolation signals "A, F, R, Z," closure time 'N/A," normal status " Closed," and remarks "From elev. 276'-6" to 27-PCA-101B. Note 13."

[cf] Containment Penetration 59: add the following valves:

. 27SOV-123E2 with isolation signals *A, F, R, Z," closure time "N/A," normal status " Closed," and remarks *From elev. 276' to 27-PCA-101 A. Note 13."

. 27SOV-123E1 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "From elev. 276' to 27-PCA-101 A Note 13."

[cg] Containment Penetration 203A: add the following valves:

. 27SOV-119E2 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "To 27-PCA-101 A. Note 13."

. 27SOV-119E1 with isolation signals "A, F, R, Z," closure time "N/A," normal status " Closed," and remarks "To 27-PCA-101 A. Note 13."

_-________w

t Attachment il to JPN-89-033 SAFETY EVALUATION Page 15 of 47 I

[ch] Containment Penetration 203B: add the following valves:

. 27SOV-124E2 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From 27-PCA-101 A. Note 13."

. 27SOV 124E1 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks "From 27-PCA-101 A. Note 13 "

. 27SOV-124F2 with isolation signals "A, F, R, Z," closure time "N/A," norrral status "Open," and remarks "From 27-PCA-1018. Note 13."

. 27SOV 124F1 with isolation signals "A, F, R, Z," closure time "N/A," normal status "Open," and remarks *From 27-PCA-101B. Note 13."

Addition of isolation Signal (s) - Category C Category C1 - OriginalPlant Design

[ci] Containment Penetrations 7A,7B,7C,7D: add isolation signals "R" and *M" for all main steam valves.

[cj] Containment Penetration 8: add isolation signals "R" and "M" to all main steam drain valves.

[ck] Containment Penetration 13A: add isolation signal *X" for valve 10MOV-25A.

[cl] Containment Penetration 138: add isolation signal "X" for valve 10MOV-25B.

Category C2 - Post-TMiModification

[cm) Containment Penetration 25 & 71: add isolation signal "H" for valves 27AOV-111 and 27AOV-112.

[cn] Containment Penetration 26A & 26B: add isolation signal "H" for valves 27AOV-113,27MOV-113, and 27AOV-114.

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Attachment 11to JPN-89G3

. SAFETY EVALUATION

.. Page 16 of 47 i ,

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[co). . Containment Penetration 205: add isolation signal "H" for the following valves:

27AOV-117 - 27MOV-117 27AOV-118 27MOV-123

. [cp] Containment Penetration 220: add isolation signal "H" for valves:

27AOV-116 27AOV-132A 27AOV-115 27AOV 132B Valve Closure Time Changes - Category D Environmental Qualification (EQ) Changes

[cq]' Containment Penetration 10: valve close time for 13MOV-15 and 13MOV-16 (RCIC steam supply) is reduced from "15" to "12" seconds.

[cr) Containment Penetration 11: valve close time for 23MOV-15 and 23MOV-16 (HPCI steam supply) is reduced from "20" to "13.5" seconds.

n

(-

_m____._._ . _ _ __ . _ _ _ __ _ _ _ _ _ _ _ _ _ _ _

Attachment 11 to JPN-89-033 !

SAFETY EVALUATION l Page 17 of 47 Section il PURPOSE OF THE PROPOSED CHANGES The overall purpose of the proposed changes to the Technical Specifications is to clarify and update Table 3.71 (" Process Pipeline Penetrating Primary Containment") and the associated notes.

These changes will update the table to accurately reflect the current condition of the plant. In !

addition, the proposed changes will result in a table format similar to the NRC's Standard Technical Specifications (STS).

Category A - General Changes l These changes clarify wording, improve nomenclature, and eliminate typographical errors which currently exist in the Technical Specifications. Table 3.7-1 was changed to describe valve function - i and identification numbers, more clearly define and describe isolation signals, and eliminate l abbreviations. !

Generalchanges to Table 3.7-1 l ltems [a] through [e]

The revised Table 3.7-1 has a new title " Primary Containment isolation Valves" (PCIV) and j format, and is organized according to containment penetration number. The proposed changes to i delete columns for " valve type", " power to open", " group", " location ref. to drywell" and " power to close" improve the table's format. The change to add pages to the table is to allow for the additionalinformation provided.

Item [f]

This change deletes the word "sec" from each close time column entry for clarification.

Item [g]

The change to add the word "(Sec)" in the close time column heading is for clarification.

Item [h]

The change of converting isolation signal "RM" to "R" is to be consistent with signals already used in the technical specifications which are a single alphabetic character. The 5,:gnal RM can be misinterpreted as signals R and M.

l

L Attachment il to JPN-89-033 SAFETY EVALUATION Page 18 of 47 Items [i], [j]

The change to use AC, DC, and N/A instead of a-c, d-c, and dashes are to be consistent with the rest of the Technical Specifications.

Item [k]

The change corrects errors in penetration numbers introduced by Amendment 48 (Reference 14). Original plant design drawings have always indicated 30a and 50c.

Item [l]

The change corrects an error in the TIP purge penetration number (35B) inadvertently introduced by a previous amendment. Onginal plant design drawings have always indicated 35E.

Item [m]

The change to delete containment penetration 36 from the table was a result of a plant modification (F178-033). The penetration was capped during the 1983 refuel outage.

' Items [n] through [x]

The change to add or delete information in the Remarks column is for clarification. Remarks that were added will assist the user of the table in determining the origin, location, or destination of the effluent in the piping system. For example, in penetration 31 Ad drywell atmosphere sample suction is taken in from elevation 276' of the drywell and transmitted to radiation monitors which are described in Note 12 of the new table.

Changes to Notes for Ta$e 3.7-1 The changes to Notes for Table 3.7-1 !s to c!arify and update the Notes section for the containment isolation valves.

Item [y]

The change to start with Note 1 rather than with the isolation signal codes is for clarification. The table of isolation signal codes has been relocated and placed at the end of Table 3.7-1.

_ _ - ___ -_A

a Attachment il to JPN-89-033

,w t SAFETY EVALUATION Page 19 of 47 Item [z] .

D The terms " suppression chamber" and ' torus" are synonymous when applied to the FitzPatrick plant.

Item [aa)

Note 4 is revised and expanded for clarification of the control rod drive hydraulic control unit isolation.

Item [ab]

Note 5 described the AC and DC electrical power sources. Since the columns " Power to Close" and

Power to Open" are deleted from Table 3.7-1, this Note is no longer necessary. The remaining 7 Notes are renumbered to account for this change.

Note 6 previously described the failure modes for each type of valve operator (motor-operated, air-operated, and solenoid-opetated). Since the column " Valve Type" is deleted from Table 3.7-1, this Note is no longer necessary.

Note 8 described valves marked in Table 3.7-1 with an asterisk (*) in the " Normal Status" column as remote manual valves. This Note is being deleted because no valves were marked with an asterisk in the table.

Note 9, previously blank, will contain the contents of Note 15. This note is unchanged.

Note 12 was deleted because it stated that closure rates were required "...for containment isolation only." This is incorrect. Some containment isolation valve closure times are limited by the environmental qualification analysis.

Note 13 was deleted because it stated that minimum valve closing time was based "...upon valve .

and line size." This is incorrect. Some containment isolation valve closure times are limited by the environmental qualification analysis.

Item lac] '

The change to renumber Note 7 is to account for notes deleted earlier in the table. In addition, this change will eliminate unnecessary abbreviations that currently exists in the Notes section.

Items [ad], [ah]

The change to renumber the Notes is to account for Notes deleted earlier in the table.

l Attachment il to JPN-89-033 SAFETY EVALUATION 1 Pege 20 of 47 ;

i Item [ae] i I

The phrase "...and is provided in this table for information only..." is added to clarify why the .

" Normal Status" column is included in Table 3.7-1. The NRC's Standard Technical Specifications j (Reference 3) does not include normal isolation valve status in the table of containment isolation valves. This note was renumbered to account for deleted notes.

~

Item [af]

The purpose of this change is to explicitly state which type of valve the Note is referring to. I Item [ag] {

i J

The change to replace the word " required" with " requires" (In Note 9) is to correct the tense. j ltem [ai] l The purpcse of this change is to explicitly state what valve the Note is referring to.

I Item [a))

The change to rearrange the sentences in the description column for signal code "A" is for clarification. There is no change to the description of the signal.

Item [ak]

This change clarifies isolation signal "D" to more clearly describe the main steam (high steam flow)line break.

l Item [al]

This change clarifies isolation signal "E" to more clearly describe the main steam (tunnel high temperature) line break.

Item [am]

The change to the description of isolation signal *G" is to clarity which of the two reactor water level signals is used.

--______n_-- __-

Attachm:nt 11 to JPN-89-033 SAFETY EVALUATION Page 21 of 47 Item [an)

The proposed change to add isolation signal "H*" (Containment High Radiation) to the isolation Signal Code Table is to reflect a plant modification necessary to fulfill the requirements of NUREG-0737 ltems ll.F.1.3 and ll.E.4.2(7). Item II.F.1.3 requires the installation of Containment High Radiation Monitors. Item II.E.4.2(7) requires closure of the containment vent and purge isolation valves on high radiation in containment.

The modification censisted of installing a High Range Radiation Monitoring System at the FitzPatrick plant to monitor high radiation in the primary containment. The installed system provides information related to the extent of core damage that may occur during an accident and will detect and measure high radiation levels within the primary containment during and following '

an accident. On receipt of a containment high radiation monitored signal, signal "H" (Containment high radiation) is transmitted to isolate containment vent and purge isolation valves. NRC acceptance of this design is documented in a January 13,1983 safety evaluation (Reference 17),

(Also see Category C2 Change items.)

Item [ao]

The change to replace the word " space" with " area" is for clarification.

Item [ap]

The word " diaphragm" is added for clarification.

Item [aq]

The addition of isolation signal *M*" (low condenser vacuum) is to reflect the as-built condition of the plant. Isolation signal "M" applies to valves associated with penetrations 7A,78,7C,7D (Main steam valves) and 8 (Main steam drain valves). This signal has always been part of the isolation signals for these penetrations. This isolation signal was inadvertently omitted during the initialissuance of Table 3.7-1.

The addition of isolation signal "X*" (Reactor vessel low water level or high drywell pressure) is to reflect the as-built condition of the plant. Isolation signal "X** applies to valves associated with penetrations 13A and 138. This signal has a! ways been part of the isolation signals for these penetrations. This isolation signal was inadvertently omitted during the initial issuance of Table 3.7-1.

Both of these signals (M and X) are part of the Primary Containment and Reactor Vessel isolation Control System (PCRVICS) and have always been tested and treated as Containment Isolation Signals.

Attachment 11 to JPN-89-033

_ SAFETY EVALUATION l .

Page 22 of 47 l .

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! Item [ar]

isolation signal "RM" is changed to "R" so that all isolation signals are a single alphabetic character; and rephrased for clarification. Signal "RM" can be misleading if read as signal *R" or "M." The description of this signal is also revised for clarity.

Item [as]

The addition of an asterisk (*) to isolation signal "U" (high reactor vessel pressure or high drywell pressure) is to reflect the original plant design. An asterisk is used in Table 3.7-1 to identify those isolation signals which are part of the Primary Containment and Reactor Vessel Isolation Control System (PCRVICS). Isolation signals not marked with an asterisk are included for information only.

This signal has always been part of the PCRVICS.

Item [at]

This change item provides information on the radiation monitors used for sampling lodine, particulate, and gaseous effluents. This will assist the user of the table in the destination and origin of the effluent specimen being tested.

Item [au]

New Note 13 describes a plant design feature which permits isolation signals "A", "F" and "Z" to be overridden. This override is necessary to permit operation of a plant modification, the addition of a Post Accident Sampling System. This modification was a direct result of the NRC's " Lesson's Learned" program (NUREG-0737, Reference 5 Item II.B.3

Post Accident Sampling" and item II.F.1.6 " Accident Monitoring - Containment Hydrogen"). This new Note applies to containment isolation valves associated with penetrations 26A,58b,58c,58d,59,203A, and 2038.

Item lav)

New Note 14 ciearly describes the containment penetrations associated with the traversing in-core probe (TIP) system. Note 14 is applicable to penetrations 35A,358,35C and 35D.

Item [aw] ;

New Note 15 clearly describes the TIP shear valves. Note 15 is applicable to penetrations 35A, i 35B, S5C and 35D.

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Attachment il to JPN-89-033 SAFETY EVALUATION

P:ge 23 of 47 a ,., .

L Category B - Addition of Valves 4

- This portion of the safety evaluation will address valves added to Table 3.7-1. This includes :

changes to reflect the original as-built condition of the plant and Post-TMI plant modifications. l l

Category B1 - Addition of Valves to Reflect Original Plant Design

.These changes reflect the as-built condition of the plant. The valves added were part of the - i original FitzPatrick plant design as described in the original FSAR (Reference 1). These valves were inadvertently omitted in the initial issuance of the Technical Specifications.- ;

l Item [ax]

i This change adds a motor operated valve to penetration 11 (High Pressure Core injection Steam l Supply). The motor operated valve was installed prior to startup. However, this change was not incorporated into either the original Final Safety Analysis Report (FSAR) or the original technical i specifications. ]

Item [ay1 This change adds two air-operated valves and two check valves to reflect the original plant l design. These valvas are shown in the original FSAR, Figure 5.2-9 (Reference 1). )

i Item [az]

This change adds air operated valve AOV-126 and CRD-138 to penetrations 37A,378,37C, & ;

37D. These valves are shown in the original FSAR, Figure 3.5-5 Sh. 2 (Reference 1).

l Item [ba]

l This change adds air operated valve AOV-127 to penetrations 38A,388,38C, & 38D. This valve is shown in the original FSAR, Figure 3.5-5 Sh. 2 (Reference 1).

Items [bb], [bc]

This change adds containment spray air test isolation valves to penetrations 39A and 398.

These valves are shown in the original FSAR, Figure 7.4-8 1

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_ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - O

Attachment ll to JPN-89-033 SAFETY EVALUATION

- Page 24 of 47 Item Ibd]

This change deletes Containment Penetration 202A from Table 3.7-1. Penetration 202A (Vacuum Breaker Reactor Building to Suppression Chamber). is .between the drywell and torus. The containment isolation valves for this penetration are associated with penetration 2028. This change is required to correct the table. See also penetration 2028.

i item [be]

Containment Penetration 202B is revised to indicate that valves 27AOV-101 A and 27AOV-101B are associated with 202B only. Penetration 202A is deleted since there is no flow in tha? line.

I Items [bf], [bg]

This change adds motor operated valves to Containment Penetration 211 A and 2118. Valves 10MOV 39A and 10MOV-39B are associated with penetrations 211 A and 211B, respectively.

Both of these valves are shown on Figure 7.4-8 of the original FSAR (Reference 1).

Item [bh]

This change adc's four containment isolation valves, three motor operated valves and one check valve, to penetration 210A. These valves are shown in the original FSAR, Figure 4.71 Sheet 1 and Figure 7.4-8 (Reference 1).

Item [bi]

This change adds three motor operated containment isolation valves to penetration 2108. These valves are shown in the original FSAR, Figures 4.71 Sh.1 and 7.4-8 (Reference 1),

item [b))

This change adds a check valve to penetration 212. The change to add the valve will complete the table of isolation valves. This valve is shown in the original FSAR, Figure 4.71 Sh.1 (Reference

- 1). This valve was inadvertently omitted during the initial issuance of the Technical Specifications.

Item [bk]

This change adds one containment isolation valve to penetrations 214. This valve is shown in j the FSAR, Figures 4.71 Sh.1 (Reference 1). The change to add the valve will complete the table of j isolation valves.

l

!' F Attachment il to JPN-89-033 SAFETY EVALUATION l- f 'J

Page 25 of 47 '

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' ltem'[b!] '

. The addition of four Containment Air Dilution (CAD) system torus purge valves to penetration {

220 is to reflect the FitzPatrick's original design. These valves are shown in the original FSAR, 1 Figure 5.2-9 (Reference 1).

Item [bm] -

This change adds penetration 221 and one check valve to the Table 3.7-2. This pr netration and valve are shown in the original FSAR, Figure 4.7-1 Sh.1 (Reference 1).

Items [bn], [bo]

This change adds four containment isolation valves to penetrations 225A and 225B in order to reflect the original plant design.

Item [bp]

This change adds condensate isolation valve 33CND-102 to penetration 228. This valve is shown in the original FSAR, Figure 10.9-1 (Reference 1).

Category B2 - Addition of Valves to Reflect Post-TMI Modifications NUREG-0737, ? Clarification of TMI Action Plan R3quirements," was issued in November 1980.

The purpose of the proposed changes in this category reflect plant modifications as part of the NRC's " Post-TMl Requirements" program.

NUREG-0737 Item II.E.4.1

Dedicated Hydrogen Penetrations" (Items [bq] and [br])

NUREG-0737, item II.E.4.1 sta+es:

Plants using external recombiners or purge systems for post accident combustible gas control of the containment atmosphere should provide containment isolation systems for external recombiner or purge systems that are dedicated to that service only, that meet the redundancy and single failure requirements of general design criteria 54 and 56 of 10 CFR 50, Appendix A, and that are sized to satisfy the flow requirements of the recombiner or purge system.

Attachment 11 to JPN-89-033 SAFETY EVALUATION

, +

Page 26 of 47 i

(1) An acceptable alternative to the dedicated penetration is a combined design that is_ both single-failure proof for containment isolation purposes and single-failure proof for operation of the recombiner or purge system.

The FitzPatrick design. incorporates a Containment Atmosphere Dilution (CAD) System which provides for control of postulated combustible gases following a Design Basis Event (DBE). The l'

CAD system is incorporated in a combined design with the Drywell and Suppression Chamber inerting and Purge system. All purging is conducted through the Standby Gas Treatment System.

The purge ventilation supply and exhaust piping to the primary containment are each provided with two fast acting air operated butterfly valves in series. Also, there originally was a bypass system with one motor operated valve around the exhaust butterfly valves from the drywell and from the torus. To insure the capability for purge initiation without requiring operator entry into the Reactor Building and to eliminate concern about a single failure while purging, a second motor operated valve was added to each of the bypass lines. This plant modification was completed during the 1981-1982 refueling outage. These motor operated valves meet the requirements of TMI-Item II.E.4.1, General Design Criteria 54 and 56 of 10 CFR 50, Appendix A, and the FitzPatrick design bases. NRC acceptance of this design was documented in a July 26,1984 letter from the NRC (Reference 18).

NUREG-0737 Item II.E.4.2 " Containment isolation Dependability" Items [bs] through [ca] RBCLCW Containment isolation Valves NUREG-0737 TMI-Item Section li.E4.2 required licensees to review operating plants for containment isolation dependability. The acceptancs criteria stated in the NUREG include General Design Criteria 54,55,56,57 and a requirement to provide diverse containment isolation signals.

In a letter dated January 7,1982, from J.P. Bayne to T.A. Ippolito (Reference 9), the Authority completed a comprehensive review of FitzPatrick's containment isolation design in response to NUREG-0737 Item II.E.4.2 - Containment isolation Dependability. This letter also includes a description of systems that are classified as essential and nonessential, and how they would be !

modified. The RBCLCW system was classified as nonesseritial.

As a result of this review, nine remote manual air operated containment isolation valves were added to nine existing lines common to the RBCLCW and ESW systems. The addition of these valves brought the RBCLCW and ESW systems into compliance with NUREG-07J7 Item II.E.4.2 -

Containment isolation Dependability. NRC acceptance of this design was documented in a NRC Inspection Report, dated August 12,1985, Reference 16. l The valves are designed in accordance with ANSI B16.34 piping as per B31.11967. The associated control electrical circuitry for the valves are classified as nuclear safety related and qualified to Seismic Class I, OA category 1, electrical class IEEE 1E, and meet the requirements of NUREG-0588

Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment."

i

Attachment il to JPN-89-033 SAFETY EVALUATION Page 27 of 47 4

These valves are equipped with a Seismic I, OA Class I air accumulator to permit up to two full cycles of valve operation in the event of loss of instrument air or nitrogen. In addition, each valve is supplied with a DC solenoid valve (IEEE Class 1E) to control the flow of instrument air or nitrogen to the valve actuator.

The valves are provided with a remote-manual isolation signal that can be activated by an operator from Panel 09-75 in the main Control Room. This configuration allows the operator to isolate these lines in the event that the Reactor Building is inaccessible.

NUREG-0737 Item II.F.1.6

Containment Hydrogen Monitor" Items ([cb] through [ch])

NUREG-0737 Item II.F.1.6 required licensees to provide a continuous indication of hydrogen concentration of the containment atmosphere in the control room. Measurement over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure was included as acceptance criteria.

The Authority installed a nuclear safety grade hydrogen monitoring system that provides the capability to detect hydrogen build up. This system continuously indicates (over a range of 0-30%)

hjdrogen concentration under both positive and negative ambient conditions.

This modification entailed the installation of Primary Containment Analyzer (PCA) sample suction and retum lines with solenoid operated containment isolation valves. The system consisted of two redundant sample trains, each capable of obtaining samples from three locations in the primary containtnent as well as the suppression chamber and secondary containment. The sample return is routed to the suppression chamber. Three of the drywell sample lines were new, using spare drywell penetrations 588,58C, and 58D. The balance of the sample lines were tapped into existing oxygen analyzer system tubing, upstream of the existing double isolation valves. These lines used penetrations 26A, 59, 203A, and 203B. Two containment isolation solenoid valves, along with appropriate integrated leak rate testing connections, are used on each sample line. The modification to add sample lines and associated containment isolation valves inects ine requirements of NUREG-0737 Item II.F.1.6, NUREG-0578, Section 2.1.9, and Reg. Guide 1.97, Rev.

2. NRC acceptance of the detign is documented in a July 25,1984 safety evaluation (Reference 15). These valves are shown on Figure 5.2-9 (Sheets 3-5) in Revision 2 of the updated FitzPatrick FSAR.

The hydrogen analyzers are located at elevation 300' of the reactor building. They are physically separated (one on each side of the containment building) and powered from separate reliable power supplies. The analyzers are designed to withstand the design basis earthquake and operate in the applicable accident environment at these locations in the reactor building. The output signal (for hydrogen concentration) from these analyzers is transmitted to the remote control cabinets in the relay room. The remote panel has alarms for high hydrogen concentration and system trouble.

All components associated with this system and associated control circuitry are nuclear safety related and are qualified to Seismic Class I, OA Category 1, and electrical Class 1 E.

Attachment 11 to JPN-89-033 l l SAFETY EVALUATION l

" I Page 28 of 47 Category C - Addition of Isolation Signals Category C1 - Addition ofIsolation Signals to Reflect Original Plant Design items [ci], [cj]

The purpose of the proposed changes to add isolation signal codes "R" and "M" to the main steam line isolation and main steam drain line isolation valves is to accurately reflect the plant's as-built condition.

The current Technical Specifications list isolation signals B,C,D,P,E for the MSIV's. This list is incomplete. Two other signals (R and M) are added to complete this list. These signals together have always been part of the FitzPatrick original design. All seven signals are part of the Primary Containment And Reactor Vessel isolation Control System (PCRVICS).

Signal "R" (which replaces "RM") added in this amendment, is a remote manual isolation signal used for all Group A & B isolation valves. This isolation function can be initiated from the control room. The original FSAR, Reference 1, Figure 7.3-4 sheet 2, shows signal code "R" (shown as RMS) for the main steam lines and main steam line drain isolation valves.

Signal "M" now represents " Low condenser vacuum-in run mode only." The Final Safety Analysis Report, Reference 2, Table 7.3-2, includes the low condenser vacuum signal as part of the isolation function for the PCRVICS. In addition, this signal is included in Technical Specifications Table 3.2-1, " Instrumentation That initiates Primary Containment isolation."

items [ck], [cl]

n the original FitzPatricie design, containment isolation valves 10MOV-25A and 10MOV-25B (penetrations 13A & 138 - RHR Shutdown Cooling and LPCI to Reactor) are required to have isolation signals R and X. However, Table 3.71 shows signal R (remote manual) as the isolation signal code. This amendment will correct this error by adding signal X (Reactor vessel low water level or high drywell pressure).

Signal X and it's description is also added to the list of isolation signal codes at the end of the table and described as follows:

X* Reactor vessel lo v water level or high drywell pressure when both shutdown cooling suction valves (10MOV-17 and 10MOV 18) are not fully closed and reactor vessel pressure is below 75 psig This signal is part of the isolation functions of the Primary Containment and Reactor Vessel isolation Control System and has always been part of the isolation signals for these valves as shown in the original FSAR, Figure 7.4-9 Sh. 2, Reference 1. 1

AttachmentiltoJPN-85 033 SAFETY EVALUATION Page 29 of 47 Category - C2 Addition of Signal H to Reflect Post TMI-Requirements (Items [cm] through [cp])

NUREG-0737 Item II.E.4.2(7) " Containment Isolation Dependability - Isolation on High Radiation" NUREG 0737 Item II.F.1.3 " Containment High Range Radiation Monitors' Item II.E.4.2(7) required closure of the containment vent and purge isolation valves on high radiation in containment. Item II.F.1.3 required installation of containment high range radiation monitors. As a result of these items, the Authority installed a High Range Radiation Monitor System at the FitzPatrick plant to monitor radiation in the primary containment after a postulated accident.

The changes also added isolation signal *H" (Containment high radiation) to containment isolation valves on penetrations 25 & 71,26A & 268,205, and 220. NRC acceptance of this design is documented in a January 3,1983 safety evaluation (Reference 17).

On receipt of a High Radiation signal, isolation signal code "H" is generated to close the vent and purge isolation valves if they are open. Signal H can be overridden by keylock switches located on the Process Control Panel. Annunciator windows are located on panel 09-4 for containment monitoring of HI-HI Radiation, Failure, and Alert.

This signal is part of the Primary Containment and Reactor Vessel isolation Control System. This system provides timely protection against the consequences of accidents involving the release of radioactive materials from the fuel and Reactor Coolant Pressure Boundary.

Category D - Change of Valve Closure Times Changes to Reflect Environmental Qualification (EQ) Program itt ems [cq] and [cr])

Tne Authority performed an EO analysis on High Energy Line B'eaks (HELB) in Reference 19.

This analysis determined the ma.dmum allowatste temperatures for equipment inside the drywell to meet EQ requirements. The changes reduces the alloweble stroke time for the inboard and outboard isolation valves on high energy lines in the Reactor Core isolation Cooling (RCIC) and High Pressure Coolant injection (HPCI) containment isolation valves. These changes are necessary to reflect assumptans in the FitzPatrick equipment qualification analyses.

The reduced valve stroke time will limit the total mass / energy release into the Reactor Duilding duririg a HELB and lower the calculated peak temperatures in the Reactor Building caused by these accidents.

I i

_________-__ - _ _____ _~

~

' Attachment ll to JPN-89-033 SAFETY EVALUATION Page 30 of 47 Section 111 - IMPACT OF THE FROPOSED CHANGES The proposed changes to the Technical Specifications will not impact plant operation or safety.

These changes clarify and update Table 3.7-1 and associated Notes as currently written in the specifications. These changes comply with current FitzPatrick design basis criteria which are discussed below.

A. Current Design Basis This portion of the safety evaluation discusses the current NRC criteria for containment isolation valves and discusses their applicability to FitzPatrick. The original FitzPatrick FSAR (Reference 1),

together with subsequent Authority commitments and NRC-imposed changes, form the current FitzPatrick design basis. The resulting criteria will be used to evaluate the changes proposed.

The NRC issued the FitzPatrick Construction Permit (CP) over eighteen years ago. The CP was issued May 20,1970 and the OL (Operating Ucense) was issued four and a half years later on October 17,1974. (The CP hearing notice was dated February 25,1970.) Most of the documents that today define NRC regulatory criteria, guidance or staff positions had not been made effective (or didn't even exist) when the CP was issued. The applicability of these documents must be established based on their effective date.

1. Original FitzPatrick FSAR The original FSAR, as supplemented, is the design basis for the FitzPatrick plant. All subsequent commitments or requirements amend this original design basis. Lacking any other requirements, FSAR governs.

The containment isolation valve design basis is defined in several sections of the original FitzPatrick FSAR. Section 1.5 " Principal Design Criteria" describes general plant design criteria.

Section 5.2.2 " Safety Design Basis" describes nine containment isolation criteria. Section 5.2.3.4

" Penetrations" describes design characteristics of pipelirne, electrical, TIP, control rod drive, personnel and equipment access penetradons. Section 5.2.3.4 " Primary Containment isolation Valves" describes the type, location and special design features of group A, B and C isolation valve.

This section alco discusses the use of automatic valve on ECCS, feedwater and control rod drive lineu. It also discusses isolation for TIP, small diameter instrument thus, and Reactor Building Closed Loop Cooling Water system. Section 5.2.4.6 " Containment Isolation" discusses how the 1 FitzPatrick design basis assures adequate protection and considers several postulated break locations. Section 5.2.5.3 " Isolation Valves" describes the valve leakage test connections, closure time testing and tests for associated coulpment.

FSAR Section 7.3 " Primary Containment and Reactor Vessel isolation Control System" describes the isolation control system. Section 7.3.2, " Safety Design Bases," lists twelve specific criterion. Table 7.3-1 " Process Piping Penetrating Primary Containment" lists the drywell penetrations. Section 7.3.4.7 " Isolation functions and settings" describes and discusses the original FitzPatrick isolation signals. Section 7.3.3 of the original FSAR defined three isolation groups: A, B and C. These groups are based on the line's service and include parts of GDC 54,55,56 and 57 of ,

Appendix A to 10 CFR 50. J

Attachment 11 to JPN-89-033 SAFETY EVALUATION

- Page 31 of 47 I 2 Applicable Regulatory Criteria l

a.10 CFR 50, Appendix A Appendix A to 10 CFR 50 " General Design Criteria for Nuclear Power Plants" was issued for comment July 11,1967. It was published again in the FEDERAL REGISTER on February 20,1971 and became effective 90 days later.

Although not published in their final form until after FitzPatrick's CP was issued, similar requirements were included in the FSAR as part of FitzPatrick's design basis. The GDC were not specifically referenced in the FSAR as design criteria. Section H.2.5 of Appendix H of the original FSAR did however compare the FitzPatrick design bases with the GDC.

" Nonessential" plant systems were upgraded to meet GDC 54,55,56 and 57 as part of NUREG-0737 ltem II.E 4.2. See section on NUREG-0737 below.

1. GeneralDesign Criterion 54 GDC 54, " Piping Systems Penetrating Primary Containment" is general and applies to all containment penetrations, regardless of service. It requires penetrations to be equipped with provisions to allow leak testing and " redundancy, reliability and performance capabilities which reflect the importance to safety of isolating these piping systems."

FSAR Section 5.2.5.3 describes two categories of leak detection. Unes open to the drywell atmosphere are provided with test connections between the two series isolation valves. This arrangement permits testing of the inboard valve if the drywell is pressurized and testing both valves if the section in-between is pressurized. A leakoff line is provided for valves that communicate with the reactor vessel. This permits leak tests of both inboard and outboard isolation valves to be conducted during reactor coolant system hydrostatic tests.

FSAR Section 7.3 describes design features to assure adequate and reliable val /e performance.

Redundancy was considered in the definition of valve groups. Groups A and B inc.ude redundant valves. In Appendix H, Section H.2.5, the FSAR stated " Provisions will also be made to dentonstrate functional performance of containment system isolation valves and permit testing of selected penetrations."

ii. GeneralDesign Criterion 55 GDC 55, " Reactor Coolant Pressure Boundary Penetrating Primary Containment," applies to lines that conr,sct directly to the reactor coolant pressure boundary. Four acceptable arrangements are described. Check valves are not acceptable as automatic valves outside containment. Other arrangements may be demonstrated acceptable; for example, instrument lines. (See discussion of Safety Guide 1.11 below.) The valves should be close to the containment and fail safe. This GDC also states that higher quality standards should be considered for these valves and lists appropriate considerations.

Attachment 11 to JPN-89-033 SAFETY EVALUATION

. Page 32 of 47 Group A isolation valves are in process lines that communicate directly with the reactor vessel and penetrate the primary containment. These lines have two isolation valves in series, one inside and one outside the primary containment.

Both the GDC and FitzPatrick FSAR include two redundant valves on this type of line.

FitzPatrick design basis criteria does not specify the valve type.

iii. GeneralDesign Criterion 56 GDC 56, " Primary Containment isolation," applies to lines that connect directly to the containment atmosphere. Four acceptable arrangements of valve type and location are described.

Other arrangements may be demonstrated acceptable; for example, instrument lines. Check valves are not acceptable as automatic valves outside containment. The valves should be close to the containment and fail safe.

The FitzPatrick FSAR defines Group B isolation valves as valves in process lines that do not communicate directly with the reactor vessel, but penetrate the primary containment and communicate with the primary containment free space. These lines generally have two isolation valves in series, both of them outside the primary containment.

Both the GDC and FitzPatrick FSAR include two redundant valves on this type of line.

FitzPatrick design basis criteria locate both valves outside primary containment but does not specify the valve type.

iv. GeneralDesign Criterion 57 GDC 57, " Closed System isolation Valves," applies to lines that are not connected to either the

! reactor coolant pressure boundary or the containment atmosphere (lines which form closed loops I

inside containment). GDC 57 requires one valve outside containment, either manual (locked closed), automatic, or remote manual. Uce of a check valve !s specifically excluded.

i Por FitzPatrick, Group C isolation vaNas are in process lines that penetrale the primary l containment but do not communicate directly with the reactor vessel, or the primary containment I free space. These lines require one isolation valve located outside the primary containment.

Remote manual or check vabes are used to isolate Group C lines at FitzPatrbk. (See original FSAR Section 5.2, page 5.2-9.)

b. Regulatory Guide 1.11 (Sarety Guide 11)

Safety Guide 11

Instrument Unes Penetrating Primary Containment,' which was issueo March 10, 1971 " describes a suitable basis...to implement General Design Criteria 55 and 56 for... Instrument lines."

A supplement to Safety Guide 11

Instrument Unes Penetrating Primary Reactor Containment l Backfitting Considerations," dated February 17, 1972, discusses the applicability of this Safety l Guide dependent on the plant's CP hearing notice date. The FitzPatrick CP hearing notice was dated February 25,1970, and the NRC staff considered the guide applicable.

Attachment 11 to JPN-89-033 SAFETY EVALUATION

. Page 33 of 47 Safety Guide 11 divides instrument lines into two classes: those instrument lines which are part of the protection system; and those not part of the protection system. Protection system lines should: a) satisfy redundancy, independence and testing requirements of protection system; b) be l sized or include an orifice to limit effects of a line break; c) be provided with an isolation valve

! outside containment, either automatic or remote manual an excess flow check is acceptable; d) l be of a quality equal to containment, testable, inspectable, and located to minimize possibility of damage; e) not increase instrument response time unacceptably. Lines not part of the protection system should meet provisions b - c above, or have two automatic isolation valves, one inside and one outside containment.

f FSAR Section 5.2.3.5 states that TIP lines and small diameter instrument lines are not provided l

with isolation valves inside the drywell. Instrument piping penetrating primary containment and connecting to the reactor coolant system is " dead-ended" at instruments in the reactor building.

These lines are provided with manual isolation valves and excess flow check valves.

c. Regulatory Guide 1.97, Rev. 2 Regulatory Guide 1.97 " Instrumentation For Ught Water Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following An Accident," Revision 2. Although not part ;f the original design basis, FitzPatrick was upgraded to meet this guide as part of its efforts to comply with Supplement No.1 to NUREG-0737. Specifically, positive valve position indication was added to many containment isolation valves.

d. Regulatory Guide 1.114 Regulatory Guide 1.114 " Containment Isolation Provisions For Fluid Systems," was issued in April 1978 and is not applicable to FitzPatrick.

e. Gtandard ReviewPlan Ponions of NUREG-0737 item II.E.4.E were basod upon U3NRC Standard R9 view Picn (NUREG-0800) Section 6.2.4 " Containment Isolation System."

f. NUREG-0737, Ism II.E.4.2 la response to NUREG-0737 item II.E.4.2, $ontainment isolation Dependability," the Authority completed a comprehensive review of the containment isolation dependability. Item II.E.4.2 required that nonessential systems comply with GDC 54,55,56 and 57, and be provided with diverse containment isolation signals.

EssentialSystems Essential systems were defined as systems which are required for, or could be of direct aid in mitigating the consequences of a postulated accident. Systems designated in the FSAR as engineered safeguard, nuclear safety or special safety systems are essential since they are specifically required for accident mitigation.

r _.

Attachment il to JPN-89433 SAFETY EVALUATION Page 34 of 47 Eleven systems have been designated " essential" for ll.E.4.2 and therefore need not comply with the GDC:

1. Control Rod Drive >

2. All modes of Residual Heat Removal (except for the . reactor head spray lines and lines which discharge to radwaste.) - i

3. Core Spray

4. High Pressure Coolant injection

5. Reactor Core isolation Cooling

6. Emergency Service Water

7. Standby Uquid Contro!

8. Drywell Instrument Air (nitrogen lines only)

9. Feedwater

10. Containment Atmosphere Dilution

11. Vacuum Relief (Reactor building to suppression chamber vacuum breakers)

NonessentialSystems l

In the event of an accident, nonessential systems penetrating primary containment are required to be automatically isolated by diverse signals in accordance with GDC 54,55,56, and 57. The following systems which penetrate primary containment have been identified as nonessential:

1. Main Steam

2. Residual Heat Removal (Reactor Head Spray Mode) l 3. Reactor Water Cleanup l- 4. Radwaste (Drywell Floor and Equipment Sump Pumps)

5. Service Air

6. Instrument Air

7. Breathing Air

8. Leak Rate Analyzers (Drywell Pressure Sensing, Torus Pressure Sensing)

9. Reactor (Recirculation Loop) Sample l

10. Suppression Chamber and Containment Atmosphere Samplir g Lines

11. Traversing incore Probes

12. Instrumentation Unes

13. Reactor Building Closed Loop Cooling Water

14. Recirculation Pump Mini-Purge

____________a

Attachment 11 to JPN-89-033 SAFETY EVALUATION Page 35 of 47 Six systems were modified to meet the requirements of GDC 54,55,56 and 57, and are actuated with diverse and redundant isolation signals as per NUREG-0737, item II.E.4.2: reactor water cleanup; traversing in-core probe; re. circulation pump mini-purge; leak rate analyzer; reactor building closed-loop cooling water; and containment vent and purge. The Emergency Service Water (ESW) system, which is part of RBCLCW, is considered essential since it has the postaccident function of providing cooling water to components of various systems required for accident mitigation. Reference 9 details the modifications and documents the study.

Category A- General Changes General changes to the table and notes are meant to clarify and revise Table 3.7-1 of the Technical Specifications. These changes are purely administrative in nature and will have no impact or effect on plant operations or safety.

Items [a] through [m]

These changes are meant to clarify and update the table and to better interpret the contents.

Items [n] through [x]

The proposed changes to the Remarks column (Category A) will have no impact on plant safety or operations since it provides the user of the table additional information regarding the process piping system.

Items [y] through [aw]

The proposed changes to Notes for Table 3.7-1 (Category A) are intended to clarify the information contained in the table and not impact plant safety or operations.

Category B - Addition of Valves Category B1 Addition of Valves to Reflect Original Plant Design (Items [ax] through [bp])

The changes in this category include corrections to reflect FitzPatrick's original design. Certain isolation valves were omitted in the initial issuance of the Final Safety Analysis Report or the Technical Specifications. These valves have been part of the containment isolation system since issuance of the Operating License. The valves added in this category have always been tested and treated like a Containment isolation Valves. Therefore, there can be no impact on plant operations or safety. These valves that are being added to the table were previously evaluated during the NRC's review of the original design of the plant.

Attschment 11 to JPN-89-033 SAFETY EVALUATION

, . Pgge 36 of 47 Category B2 Addition of Valves to Reflect Pout-TMI Requirements Modification Program The addition of valves to comply with the Post-TMI Requirement program v2 not impact plant operations, because the modifications provide additional assurance that the isolation valves will perform their design function and meet all applicable containment isolation criteria.

NUREG-0737 Item II.E.4.1 " Dedicated Hydrogen Penetrations" (Items [bq] and [br])

The vent and purge containment isolation bypass valves were installed to comply with TMI-Item II.E.4.1, GDC 54 and 56 of Appendix A to 10 CFR 50, and the FitzPatrick design bases. This modification has improved overall plant safety by: 1) meeting redundancy and single-failure requirements for containment isolation and 2) providing fail-safe logic to the individual override controls for the purge valves. NRC acceptance of this design was documented in a July 26,1984 letter from the NRC (Reference 18).

l The vent and purge system (which is part of the Containment Atmosphere Dilution System) is I

not used in the beginning of an accident. This system is normally isolated during reactor operation and, even if it was open, it automatically isolates on receipt of a containment isolation signal. The system remains isolated until the operator has determined the need for its use. At the appropriate point in the accident sequence the operator must take positive action to open these valves.

This system can be used to vent the containment atmosphere to the stack, if desired (Reference 2). This line is routed to the Standby Gas Treatment System so that release of gases from the primary containment is controlled.

The added vent and purge containment isolation valves will improve the systems's capability to I

reliably isolate in the event of a postulated accident, thereby increasing the reliability of containment integrity. Spurious actuation of these containment isolation valves will not result in a plant transient; nor will it affect the plant's safe shutdown capability.

Design features of the containment isolatinn valves, such as redundant flow-paths, will assure that the vent and purge system can perform both its containment isolation function and post-accident function assuming any postu!ated single-failure.

l l

The logic for operation of the vent and purge bypass containment isolation valves is fail-sate. )

This logio is powered from both AC and DC sources such that no single failure can prevent purging or cause loss of containment integrity from either the drywell or the torus. This provides a fail-safe purge path from the containment under postulated accident conditions.

The isolation signals to the vent and purge containment isolation valves can be overridden when the CAD system is manually initiated. This is accomplished by keylock switches at the purge panel located in the relay room. This override feature allows the valves to be manually opened with an isolation signal present and yet automatically close if a diverse isolation signal is generated.

l l

l

F Attachment il to JPN-89-033 l b SAFETY EVAWATION Page 37 of 47' FSAR Section 5.2.3.5 " Primary Containment isolation Valves" states:

Valves on lines that communicate with the primary containment free space require two valves in series, both located outside the primary containment and as close to the primary containment boundary as practical.... Both valves close automatically on a group 11 isolation signal and receive the isolation closure during reactor operation.

Isolation of the existing vent and purge bypass valves isolate on the following group 11 signals:

A- reactor vessellow water level F- high drywell pressure H - containment high radiation R - remote manual switch from' control room Z - reactor building ventilation exhaust high radiation The_ motor-operated valves on the bypass lines actuate on the same isolation signals as the existing vent and purge valves. The addition of signal "H" is discussed in Category C changes.

NUREG-0737 Item IIE.4.2 " Containment Isolation Dependability" (Items [bs] through lca])

The RBCLCW lines do not communicate with the reactor coolant system or the containment atmosphere. The RBCLCW piping serves as the first barrier for containing postulated radioactive releases. Prior to the installation of these new containment isolation valves, this was the primary barrier for containment isolation. The manual isolation valves outside the containment were not considered a second barrier for containment isolation because these valves are not accessible during a postulated accident, and thus no credit was taken for these valves as a containment isolation barrier.

The installation of the power-operated valves in the RBCLCW lines has improved overall plant safety by increasing containment isolation dependab9ity. The air-operaMd remcae manual icolation valves serve as the second barrier for containment isolation. These two barriers meet the redundancy requirement for containment isolation as dened in GDC 57.

The addition of these new containment isolation valves will not inhibit the operation of the RBCLCW or ESW systems. Spurious iso!ation of the RBCLCW lines during normal operation may result in a plant trip due to the inability to cool equipment located in the drywell. During accident conditions these lines allow the use of ESW to remove heat from the containment. Therefore, it is preferred the valves fail-open upon loss of instrument air or nitrogen and DC power supply. This assures an uninterrupted cooling water supply during normal conditions and allows Emergency Service Water to supply drywell equipment with cooling service water during accident conditions.

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Attrchment 11 to JPN-89-033 SAFETY EVALUATION

, Page 38 of 47 Each isolation valve is equipped with a Seismic I, OA Class I air accumulator to permit up to two full cycles of valve operation in the event of loss of instrument air or nitrogen. In addition, each valve is supplied with a DC solenoid valve (IEEE Class 1E) to control the flow of instrument air or nitrogen to the valve actuator.

The valve internals are qualified to the containment accident conditions. The valve operator, and equipment are qualified to the reactor building environment. The cable and conduit and electrical equipment are designed in accordance with 10 CFR 50, Appendix R, Fire Protection requirements.

The valves are provided with a remote-manual isolation signal that can be activated by an operator from Panel 09-75 in the rnain control room. This configuration allows the operator to isolate these lines in the event that the Reactor Building is inaccessible.

NUREG-0737 Item II.F.1.6 " Containment Hydrogen Monitor System" (Items [cb] through [ch])

The hydrogen monitoring system provides the capability to determine the hydrogen concentration inside the containment after an accident. The added containment hydrogen monitoring system lines and corresponding containment isolation valves will improve containment isolation and hydrogen monitoring. This system is a nonessential system since it is not required to mitigate the consequences of an accident. In the event of an accident, this system is required to be automatically isolated by diverse signals.

The sample lines penetrate primary containment and are open to the containment free space. In accordance with Sections 5.2.3.5 and 7.3.3 of the FitzPatrick FSAR, these lines are classified as Group B. Group B isolation valves are valves in process lines that do not communicate directly with the reactor vessel, but penetrate the primary containment and communicate with the free space.

These lines have two isolation valves in series, both of them outside the primary containment. The added containment isolation valves are arranged in series; two valves located outside of containment for each samp!e line. The valves fail-ctose by spririg force and therefore do not require redundant e:ectrical powar supplies. Both valves on each line are powered from the same sourca.

This arrangement aesures that the fa!!ure of one power source will not disab'e both sample lines.

Spurious actuation of tha containment isolation valves will result in padial loss of H2 /02 indication. This will not result in a plant vansient or aNect the plent's shutdown capatility.

Certain hydrogen monitoring conta!nment isolation valves are normally open. Inadvertent closure of these during ar> accident has limited effects, sirce the containment hydrogen monitoring system is a nonessential system. The valve posinons are indicated and alarmed in the control room.

All valves in this system are powered from the emergency buses and, therefore, can be opened after a LOCA, loss-of-offsite-power, or any single-failute of a powe supply.

1

Attachment il to JPN-89-033 SAFETY EVALUATION Page 39 of 47 The new valves are isolated from the same signals which closed the original valves. These signals are:

A- reactor vessellow waterlevel F- high drywell pressure R - remote manual switch from control room Z - reactor building ventilation exhaust high radiation The new valves satisfy the FSAR isolation criteria, since they serve the same function and are controlled from the same isolation signals.

This modification enhances plant safety by increasing containment isolation dependability and hydrogen monitoring. No different failure modes are created; nor is the safety function of the hydrogen monitoring system changed by the addition of these containment isolation valves.

Category C - Addition of isolation Signals

\ Category C1 Addition of Signals To Reflect Original Plant Design l

l l

l Items [ci], [cj]

l l

The addition of signals R & M to the main steam line and main steam line drain valves (Containment Penetrations 7A,78,7C,7D & 8) completes the list of isolation signal codes for these valves.

l The seven signals (B,C,D,E,M,P,R) toge'.her are part of the isolation functions of the Primary L Containment And Reactor Vessel ! solation Controf System (PCRVICS). The signal code "R" which replaces "RM", in this proposed amendment, is a remote manual isolation code signal used for all Group A & B isolation valves. This isolation function can t'e controileo from the reactor control I room.

The addition of solation signal code "M" (Low Gonderwar Vacuum while in the run mode) is for i clarification. The Final Safety Analysis Report, Reference 2, includes this signal as part of the l isolation function for the PCRVICS.

This change improves the table to ensure consistency throughout the Technical Specifications.

1 items [ck], [cl]

l l The addition of signal X for containment isolation valves 10MOV-25A and 10MOV-25B (penetration 13A & 138 - Shutdown Cooling and LPCI to Reactor) will not impact plant safety or operations. This signal code was omitted from the table during the development and preparation of the original issuance of the Technical Specifications. This signal is being added to make Table 3.7-1 complete and consistent. The Authority's records indicate that isolation X was installed on these valves as part of the original FitzPatrick design.

Signal X is part of the isolation functions of the PCRVICS.

Attachment 11 to JPN-89-033 SAFETY EVALUATION ;

P:ge 40 of 47 )

Category C2 Addition of Signal H to Reflect Post TMI-Requirements Program Modification items [cm] through [cp] j I

The primary containment system, as stated in FSAR Section 5.2.1, provides the capability, in l conjunction with other engineering safeguards, to limit the release of fission products in the event of postulated accidents. The addition of isolation signal "H" (Containment high radiation) to the vent and purge system increases the primary safety objective. This wi!. assure that containment isolates during accidents which could release a significant amount of radioactivity to the containment atmosphere. This prohibits the release of airborne radioactive materials from the containment.

These signals are diverse and redundant to assure dependable isolation.

The trip setpoint for isolation signal H is set such that it will not inadvertently isolate Primary Containment during normal plant operations. Therefore, there can be no impact on plant operations or safety. The probability that automatic isolation on high radiation would be required is extremely low. Several diverse methods already exist for the detection of primary coolant leakage that could indicate to the operator that a high radiation condition exist.

The range and sensitivity of the radiation monitors used for sensing the radioactivity in the drywell is sufficient to assure timely closure of the vent and purge valves.

This signalis also part of the Primary Containment and Reactor Vessel isolation Control System.

This system assures timely protection against the consequences of accidents involving the release of radioactive materials from the fuel and Reactor Coolant Pressure Boundary.

Categcry D - Change of Valve Closure Times (items [cq) and [cr])

Environmental Qualification Program (EQ)

These changes were made to reduce potential environmental effects of a high energy line break (HELB) in the Resctor Cc,re lsolatlori Cooling (RCIC), and High Pressure Coolant injection (HPCI) systems Thesc changes dacrease the maximum allowable stroke time for the inboard and outboard isolation valves.

Faster valve closure time decreases the total mass / energy released into the Reactor Building and consequently lowers the peak temperature in the Reactor Buildino caused by a HELB. j The containment isolation valves for the HPCI and HCIC System steam supply line from tha reactor vessel are normally open during plant operat;on.

The decreased stroke time for these valves will have no effect on system capability. For plant conditions when RCIC is automatically isolated, the turbine trip valve will close on the same signals as the isolation valves. The turbine trip valve will close in a fraction of the time required to close the isolation valves. Therefore, the decreased stroke time for the RCIC system high energy lines will have no effect on RCIC system automatic isolation.

These changes will improve plant safety, by limiting the effects of a HELB on nearby equipment, without impact on normal operation or the criteria set forth by the FitzPatrick FSAR.

Attachment il to JPN-89-033 SAFETY EVALUATION

. Page 41 of 47 Section IV EVALUATION OF SIGNIFICANT HAZARDOUS CONSIDERATIONS The proposed changes to the James A. FitzPatrick Technical Specifications involve no significant hazard considerations as defined in 10 CFR 50.92. They are changes such as: correction of an error; a change in nomenclature; clarification of a specification; or involve modifications mandated by the NRC. Operation of the FitzPatrick plant in accordance with the proposed amendment would not involve significant hazards considerations .

The following paragraphs describe, for each of the four categories defined, why no significant hazards consideration is involved.

Categories A, B1, & C1 Changes to Table 3.7-1 and Notes:

(1) 'will not significantly increase the probability or consequences of an accident previously evaluated because the changes: clarify the table; provide better nomenclature; and eliminate typographical errors. These changes are purely administrative in nature and do not involve changes to procedures or facility modifications and cannot increase the probability of an accident previously evaluated.

(2) will not create the possibility of a new or different kind of accident previously evaluated because the proposed general changes are administrative in nature and improve the description and function of the containment penetration table and associated Notes.

(3) will not involve a significant reduction in the margin of safety because the changes are purely administrative. Valves that have been added to the table which have always been part of the FitzPatrick's original design were reviewed by the NRC in the original FSAR submittal. These valves have always been tested and treated as containment isolation valves.

l Category B2 Changes renarding the addition of valves to Table 3.7-1 to reflect TMI Modifications:

(1) will not significantly incre&se the probability or consequences of an accident previously evaluated.

Il2.4.1 Dedicated Hydrogen Penetrations The added vent and purge containment isolation valves will improve the systems's capability to reliably isolate in the event of a postulated accident, thereb" increasing the reliability of containment integrity. The vent and purge system is an essential system and is not used in the beginning of a design bases LOCA. The system is normally isolated and l remains isolated until the operator has determined the need for its use.

Design features of the containment isolation valves, such as redundant flow-paths, will assure that the vent and purge system can perform both its containment isolation function and post-accident function assuming any postulated single-failure.

Attachment il to JPN-89-033 SAFETY EVALUATION

., i Page 42 of 47 The logic for operation of the vent and purge bypass valves is fail-safe. This logic is -

powered from both AC and DC sources such that no single-failure can prevent purging or cause loss of containment integrity from either the drywell or the torus. This provides a fail-safe purge path from the containment under postulated accident conditions.

II.E.4.2. Containment isolation Dependability The added RBCLCW containment isolation valves improve system containment isolation dependability. The RBCLCW system is an essential system in the Emergency Service Water mode. The valves are designed to fail-open to assure operation of RBCLCW and ESW systems. The systems safety function is unchanged by this modification.

II.F.1.6 Containment Hydrogen Monitoring The added containment isolation valves in the hydrogen monitoring system will improve

! containment isolation dependability. The modification also provided an alternate flowpath to i assure continuous H 2 monitoring capability. The containment isolation valves meet the redundancy and single-failure requirements for containment isolation.

(2) will not create the possibility of a new or different kind of accident previously evaluated.

II.E.4.1 Dedicated Hydrogen Penetrations The added vent and purge containment isolation valves will improve the systems's capability to reliably isolate in the event of a postulated accident. Spurious actuation of these corctainment isolation valves wiM not result in a piant transiertt; nor will it affect the

, plant's safe shutdown capability.

L N ll.E.4.2 ContainrosntIsolation Dependability The added RBCLCW containment isolation valves will improve system containment isolation dependability. Spurious isolason of these valves and the' concomitant loss of drywel!

L cooling is precludeo by design features such as iail-opening of the valves on loss of l

instrument air or nitrogen. This will allow continuous operation of the RBCLCW and ESW i

systems. No new failure modes ace created; nor isthe safety function of the RBCLCW and ESW systems changed by the addition of these containment isolation valves.

l l

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Attachment 11 to JPN-89-033 SAFETY EVALUATION i

Page 43 of 47

> 1 1

II.F.1.6 Containment Hydrogen M:nitoring The added containment hydrogen monitoring system lines and corresponding containment isolation valves improve hydrogen monitoring capability and containment isolation.

Spurious actuation of these containment isolation valves will result in the partial loss of H 2

/02 indication. This will not result in a plant transient nor affect the plant's safe shutdown capability.

The added containment isolation valves are arranged in series; two valves located outside of containment for each sample line. Both valves on each line are powered from the same source. This arrangement assures that the failure of one power source will not disable both sample lines. Spurious opening of these valves is precluded by design features such as fail-close by spring force. No different failure modes are created; nor is the safety function of the hydrogen monitoring system changed by the addition of these containment isolation valves.

(3) will not involve a significant reduction in the margin of safety.

II.E.4.1 Dedicated Hydrogen Penetrations The modification to the vent and purge system does not reduce the margin of safety because 1) the modification improves upon redundancy and single-failure requirements for containment isolation; and 2) the logic to the individual override controls is fail-safe for the vent / purge valves. NRC acceptance of this design was documented in a July 26,1984 letter from the NRC (Reference 18). l II.E.4.2 Containment Isolation Dependability The added RBCLCW containment isolation valves will improve safety by increasing containment notation dependability. The RBCLCW piping !s the first barrier for contair;ing postulated radioactive releases. Prior to the installation cd these new containment isolation valves, this was the primary barrier for containment isolat:on. The safety function of the RBCtCW system is not changed by tha addition of these containment isolatior; valves.

Design features such es fail-opening of the valves will assure system operation dunng cod after an accident.

II.F.1.6 Co.iainment Hydrogen Monitoring The added containment hydrogen monitor system lines and corresponding isolation valves do not reduce the margin of safety because 1) containment isolation meets single-failure requirements through fail-safe logic; and 2) hydrogen monitoring has redundant flowpaths.

4

Attachment 11 t)JPN-89-033 SAFETY EVALUATION Page 44 of 47 Category C Changes regarding the Addition ofIsolation Signals:

(1) will not involve a significant increase in the probability or consequences of an accident previously evaluated.

C1 changes regarding the addition of signals to reflect originalpiant design isolation codes "M", "R", and "X" are not new isolation signals for the containment isolation valves, since they were part of FitzPatrick's original design. They were reviewed by the NRC in the original FSAR submittal. These isolation signal codes have always been tested and treated as part of the plants protection system.

C2 changes regarding the addition of signal "H" to reflect TMI-Item II.E.4.2(7)

" Containment Isolation Dependability - Isolation on High Radiation" The addition of signal "H" (Containment high radiation) to the vent and purge containment isolation valves improve the systems's capability to isolate in the event of high radiation in the drywell. This new isolation signal increases the redundancy and diversity of physical parameters sensed to isolate the containment in the event of an accident.

(2) will not create the possibility of a new or different kind of accident.

C1 changes regarding the addition of Mgnals to reflect originalplant design Signals "M","R", and "X" are added to Table 3.7-1 to complete the list for certain containment penetrations in order to rnake the Technical Specifications reflect the original plant design. The addition of these signals are purely administrative in nature and were .

reviewed by the NRC in the original FSAR submittal.

C2 changes regarding the addition of signal "H" to reflect TMI-Item II.E.4.2(7)

"Containnmnt Isolation Dependability - Isolation on High Radiation" The addition of isolation signal "H" irnproves diversity of physical paramaters sensed to isolate the containment in the event of an accidcnt. The trio setpoint for this signal is set '

such that it will not spuriously isolate the vent and purge system when the valves are being l used.

(3) will not involve a significant reduction in the margin of safety. 1 C1 changes regarding the addition of signals to reflect originalplent design ,

l These changes have always been part of the original design of the plant as reviewed by the NRC and documented in the design basis of the FSAR. The changes are administrative in nature and correct the list of signal codes to reflect FitzPatrick's original design. ,

I i

1

_ _ _____ _ _ i

1 Attachment il to JPN-89-033 ;

f SAFETY EVALUATION ;

. Page 45 of 47 C2 changes regarding the addition of signal "H" -

to reflect TMI-Item 11 2 . 4 . 2 ( 7) l

" Containment Isolation Dependability - Isolation on High Radiation" l This modification increases safety because of the added ability to sense high drywell'.

radiation and isolate primary containment after a postulated accident. The addition of signal "H" (containment high radiation) increases the redundancy and diversity of physical parameters sensed to isolate the containment in the event of an accident. This provides additional assurance that releases during an accident are minimized. i l

Category D Changes regarding Valve Closure Time i (1) will not involve a significant increase in the probability or consequences of an accident previously evaluated because the decreased stroke time is still within the valves and valve !

operators design capabilities. Decreased valve closure time will not significantly affect the l design or operational characteristics of these systems. j (2) will not create the possibility of a new or different kind of accident because no new failure l modes are introduced that will affect the operation of Reactor Core Isolation Cooling and i High Pressure Coolant injection systems.

(3) will not involve a significant reduction in the margin of safety. These changes will increase safety by ensuring that area installed equipment will operate in environmentally qualified conditions that are postulated in the event of an accident.

Similar Examples .

1 t'n the April 61983 FEDERAL REGISTER (48FR14870), the NRC published examples of licensed Amendments that are considered not likely to involve significant hazards considerations. Example number (i) of that list is applicable to proposed administrative changes change (Ittm3 [a] through

[bp] and [ci] through [ci]) and states.

1 "A purely administrative change to the technical specifications: for example a change to achievo consistency throughout the technical specifications, correction of an error, ora change in nevnenclature." l I

Proposed change items [bq) through [ch] and [cm) through [cr] are similar to example (vii) which l states:

"A change to make a license conform to changes in regulations, where the license results in very minor changes to facility operations clearlyin keeping with the regulations."

L I- I '

1

_ _ _ - _ ______________-___-_____a

f Attachment ll tsJPN-89-033 SAFETY EVALUATION

, i Page 46 of 47 e

Section V IMPLEMENTATION OF THE PROPOSED CHANGES

. Implementation of these changes, as proposed, will not impact the ALARA, Security, or Fire Protection Programs at FitzPatrick, nor wll1 the changes impact the environment.

Section VI CONCLUSION .

V The proposed changes do not constitute an unreviewed safety question as defined in 10 CFR 50.59. The changes:

a. will not change the probability or 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 type different from any previously evaluated in the Safety Analysis Report;

c. will not reduce the margin of safety as defined in the basis for any technical specification;

d. Jo not constitute an unreviewed safety question; and

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

Section Vil REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Original Final Safety Analysis Report (FSAR as submitted with license application).

Figure 3.5-5 Sh. 2 " Flow Diagram-Control Rod Drive System" Figure 4.7-1 Sh.1 " Reactor Core isolation Cooling System" Figure 5.2-9 "Drywell and Suppression Chamber inerting and Purge System" Figure 7.3-4 Sh. 2 " Reactor Coolant System"

'- Figure 7.4-1 Sh.1 "High Pressure Coolant injection System" Figure 7A 8 " Residual Heat Removal Gystem" Figure 7.4-9 "Besidual Heat Removal System Sh.1 a 2 (FCD)"

Figure 10.9-1

Condensate System"

2. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analys:s Report (FSAR)

Sections: 4.9, " Reactor Coolant System Design Criteria"; 5.2, " Containment Design Criteria";

l 8, " Electrical Design Critena"). Tables: 522, " Primary Containment System and Associated l

Isolation Valves"; 7.3-1, " Process Pipeline Penetrating Primary Containment." Figure 4.9-1,

" Reactor Water Cleanup System - Piping end instrumentation Diagrarn. Sheet 1 of 2."

3. NUREG-0123, " Standard Technical Specifications for General Electric Boiling Water Reactors", Revision 3 (1980).

4. NUREG-0578, "TMI-2 Lessons learned Task Force Status Report and Short Term l

Recommendations", published July 1979.

l 1'

Attachment 11 to JPN-89-033 SAFETY EVALUATION

, . Page 47 of 47 y.

Section Vil REFERENCES (con't)

5. NUREG-0737, " Clarification of TMI Action Plan Requirements," published November 1980.

6. NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition, Section 6.2.4," Containment isolation System.

7. Regulatory Guide 1.97 " Instrumentation For Ught Water Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following An Accident," Revision 2.

8. PASNY letter, R. J. Pasternak to B. H. Gaer, dated August 24,1981 (JAFP-81-0871) regarding NRC I&E Inspection Report 81-07. Commits to install Reactor Building Closed Loop Cooling Water containment isolation valves.

9. PASNY letter, J. P. Bayne to T. A. Ippolitto, dated January 7,1982 (JPN-82-5) regarding NUREG-0737 Item II.E.4.2 " Containment isolation Dependability."

10. PASNY letter, J. P. Bayne to T. A. Ippolitto, dated February 22,1982 (JPN-82-21) regarding NUREG-0737 Item II.E.4.2 " Containment Isolation Dependability."

11. NYPA letter, J. P. Bayne to D. B. Vassallo, dated June 15,1984 (JPN-84-37) confirms that the containment penetrations associated with the purge system meet General Design Criteria 54 and 56 of Appendix A to 10 CFR 50.

12. " Safety Evaluat!on by the Directorate of Ucensing U. S. Atomic Energy Commission in the matter of the Power Authority of the State of New York - Jamet A. FitzPatrick Nuclear Power Plant - Docket No. 50-333," dated November 20,1972.

13. " Supplement No.1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant -

Docket No. 50-333," U. S. Atomic Energy Commission - Directorate of Ucensing, dated February 1,1973.

14. NRC letter, T. A. Ippolitto to G. T. Berry, dated January 23,1980 issued Amendment No. 48 to the James A. FitzPatrick Nuclear Power Plant Operating Ucense Technical Specifications.

Adds or revises specification for instrument requirements as approved by Amendments 8,14 and 40. Also corrects errors and inconsistencies.

15. NRC letter, D. B. Vassallo to J. P. Bayne, dated July 25,1984 regarding NUREG-0737 Item II.F.1.6 Resolved - Containment Hydrogen Monitor.

16. NRC letter, S. Collins to H. Glovier, dated August 12,1985, regarding closure of inspection item 81-07-01 (RBCLCW containment isolation valves).

17. NRC letter, D. B. Vassallo to J. P. Bayne dated January 13, 1983 concerning the Safety Evaluation on NUREG-0737 items ll.E.4.2(7) " Radiation Signal on Purge Valves" and ll.F.1.3

" Containment High Radiation Monitors."

18. NRC letter, D. B. Vassallo to J. P. Bayne, dated July 26,1984 regarding NUREG-0737 Item II.E.4.1 Resolved - Dedicated penetrations for extemal recombiner or post accident external purge system.

19. Stone & Webster letter, dated July 1,1983, regarding James A. FitzPatrick High Energy Une Break Analysis.

l l

_.____.-D

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