ML19257F050
| ML19257F050 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/01/1980 |
| From: | Cox T Office of Nuclear Reactor Regulation |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 IEB-79-05C, IEB-79-5C, NUDOCS 8007030491 | |
| Download: ML19257F050 (10) | |
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...e NOTE T0: John F. Stolz FROM:
Tom Cox
SUBJECT:
TMI-1 RESTART SER - TECHNICAL AUDIT As we discussed on Tuesday 4/29, this note is to summarize my recent efforts in the subject review. Based on the list of issues (Enclosure 1) given re by Harley Silver, I looked at four items that Harley felt were technically complex er.ough to benefit from a review by an LPM. The material reviewed was the Status Report of January 11,1980 and a draft supplement (dated 4/21/S0) to the Status Report.
I did not attempt to review the documents in their entirety to determine whether all matters in the NRC Action Plan (NUREG-0560) were addressed. Harley has said that the Status Report and Supplement have been written to address only the requirements of the NRC Order and Notice of Hearing of 08/09/79. The SER probably should address, as a minimum, each of the items categorized as "Ful? Power Requirements" in a draft paper available from Warren Minners. This paper describes and categorizes the specific TMI-related actions for new operating licenses as contained in the Action Plan.
The items I reviewed were:
Order Item 2, Bulletin 79-05C, items -2 & -3.
Order Item 8, Items 2.1.4, 2.1.7.a. and 2.1.7.b.
Order Item 2 required NRC review and approval of "all applicable actions speci-fied in IE Bulletins79-05A, 79-05B, and 79-05C." Items -2 and -3 of 79-05C involve additional SBLOCA analyses, both generic and plant-specific, and appro-priately revised operator guidelines.
It appears that the SER material available is adequate in content, and my effort here was only to combine and rewrite the material in the Status Report (page C2-16) and the supplement (page 2-2),
which rewrite is enclosed (Enclosure 2).
Regarding Order Item 8, Item 2.1.4,_I felt that our statement of position on containment isolation is reasonably clear, but that our justification for accepting the licensee's taking several exceptions to our position should be carefully reviewed by DSI and DOL management levels prior to comitting the NRC to a position to be taken at THI-1 restart hearings. I have combined and rewritten the available information to more effectively present our current position (Enclosure 3). But I recommend that our technical management be fully aware of and prepared to defend the correctness of our position accepting the licensee's argument to place more importance on assuring the capability to open the reactor coolant pressure boundary (letdown line) for post accident,relatively long tenn cooling purposes than on the capability to automatically close the pressure' boundary in the event of a LOCA.
Further, the licensee argued for exemption of the reactor coolant pump motor cooling water lines from the auto-isolation requirement on the basis that this water supply is " essential" to
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Joim F. Stoiz j continued pump operation post-LOCA. This seems cuestionable given that the pumps cr cc: rate "indefi 4* '7" d"~?t ratar coeling water as long as seal-injection water, a safety-grade system, is available.
Another mctter for recons'.co :tica is tht t ' : ;taff position as expressed in 4 points on page C 8-21 of the 1/11/E0 Stctus Report does not refer to the need for a high radiation containment isolation signal for those lines providing an "open path from the contair. cent to the environs." This requirement has been expressed both in SRP 6.2.a, Rev. 1 (page 6.2.d-5, May 1978) and in the Action Plan Section II.E.4.2.
While the position expressed in the TMI-1 SER includes the statement that " designs shall comply with the recommendations of NRC SRP 6.2.4,".there is no further discussion of the necessity for a containment radiation isolation signal.
Regarding Items 2.1.7.a and 2.1.7.b, I believe our acceptance of the proposed designs is warranted, but have some specific recommendations on the SER writeup to more accurately tnd completely express our conclusions and justification for those conclusicas. These are incorporated in Enclosure 4.
Thomas H. Cox, Sr. Project Manager Safety Program Evaluation Branch cc:
H. Silver N. Wagner
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- ENCLOSUE 2 IE BULLETIN 79-05C - SHORT TERM "1.
In the interim, until the design change required by the long-term action of this Bulletin has been incorporated, institute the follcuing actions at your facilities:
a.
Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating RCP's.
b.
Provide two licensed operators in the control room at all times during operation to accomplish this action and other imediate and followup actions required during such an occurence. For facilities with dual control rooms, a total of three licensed operators in the dual control room at all times meets the requirements of this Bulletin."
The licensee has initiated revisions to three procedures to require an imediate trip of all operating RCP's upon initiation of HPI caused by low RCS pressure.
These procedures are: EP 1202-6, Loss of Reactor Coolant / Reactor Coolant Pressure; EP 1202-5, OTSG Tube Rupture; and AB 1203-24, Steam Supply System Rupture. To ensure the actions required by these procedures can be accomplished, the licensee has stated that 2 licensed operators will be stationed in the control room at all times wnen the reactor is critical.
The licensee has committed to meet this part of the Bulletin. Full compliance will be verified when the TMI-1 procedures have been satisfactorily revised.
"2.
Perform and submit a report of LOCA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip. For each pair of values of the parameters, determine the peak cladding temperature (PCT) which results. The range of values for each parameter must be wide enough to assure that the maximum PCT or, if appro-priate, the region contsining PCT's greater than 2200 degrees F is identified."
"3.
Based on the analyses done under Item 2 above, develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements. For Babcock & Wilcox designed reactors, such guidelines should include appropriste requirements to fill the steam generators to a higher level, following RCP trip, to promote natural circulation flow."
In November IU9, Babcock & Wilcox submitted generic data entitled "Small Break Operating Guidelines," and supporting analyses for these guidelines. We have required each operating reactor licensee to submit a plant specific evaluation of these guidelines and analyses, as the response to Items 2 and 3 of Bulletin 79-05C. The TliI-1 licensee has submitted such material for our review and approval, in Amendment 15 to the licensee's Restart Report. Among other appropriate and acceptable material, the licensee's new guidelines include the requirement for an imediate RCP trip for a small break LOCA, in accordance with our findings in NUREG-0623, " Generic Assessment of Delayed Reactor Coolant
. (Enclosure 2)
Pump trip t.rir.; Small Break L ;s-.T-Coolcnt t.:;idsat2 in Pressurized Water Reactors. " With regard to Item 2, then, we conclude that, subject to final verification of the licensee's analyses, this item has been completed in compliance with the Order.
The licensee's guidelines for small-break loss-of-coolant emergency procedures includes both immediate action to trip the RCP's and to assure filling the steam generaters to enhance the begining of naturcl circulation through the reactor coolant system. While our review of the guidelines is not complete, we can conclude that the requirements of item 3 of the Bulletin 79-050 have been completed in compliance with the Order.
_.00 0
E:CLCSURE 3 2.1.4 Containment Isolation Position 1
2 3
4 Discussion The licensee has submitted descriptions and analyses in the Startup Report, as amended, to demonstrate compliance with our requirements as expressed in the above Position Statements.
Compliance with those position statements will be discussed for each of several systems.
Diversity in the sensors used to initiate containment isolation has been reported by the licensee. System variables used which by our standards qualify as diverse parameters are the reactor coolant system pressure (as used in the reactor protection system, RPS) and the containment pressure (as used in the engineered safety features actuation system, ESFAS). The licensee proposes to use both the containment high pressure (4 psig) from the ESFAS and the RPS trip signal penetrating containment. /or" logic to isolate all non-essential systemsNon-esse (1900 psig) in an "either part of ECCS operation or reactor coolant pump operation. There is inherent redundancy provided for these diverse signals, in that the ESFAS system senses containment pressure and reactor coolant system pressure in several sensor channels, and the RPS system of course redundantly senses reactor coolant system pressure, among other system variables.,
Use of the RPS system coolant. pressure trip at 1900 psig (or other RPS s1]nals) to initiate containment isolation assures such isolation at a high enough RC pressure to avoid radioactive releases for several events for which reactor coolant pressure might not even decrease to the ESFAS trip setpoint (1600 psig).
These events for which containment isolation will be initiated by the RPS trip setpoint include:
d.
Rod withdrawal accidents
- b. ~ Loss of coolant flow c.
Feedwater line break or loss of feedwater d.
Small Steam line break accident outside co~ntainrent (isolation of contain-ment lines is still desirable)
S O
2-(Enclosure 3) e.
Ejected red accident f.
Boron dilution accident g.
Cold water addition h.
Iodine spikes or crud burst after trip
- i. Loss of offsite power or station blackout.
1 We ccmcur with the licensee that the RPS t*ip signal and the containment high pressure (4 psig) signal are acceptable diverse parameters to initiate auto-matic containment isolation. Detailed systems identificr. tion as " essential" or "non-essential" has (has not) been provided by the licensee in Amendment to the Startup Report. We concur (do not concur) in the assignment of these systems.
One set of coolant lines penetrating containment are automatically isolated in a::crdance with slightly different criteria. These are the cooling water supply and return lines to the nomal air coolers in containment. They are isciated on either containment high pressure or the ESFAS signal (4 psig con-tainment pressure or 1600 psig reactor coolant pressure). The air cooler water lines are not part of the RCPB and do not open to the containment atmosphere.
Our judgement is that there is little additional risk involved in isolating these penetrations at the lower reactor coolant pressure of the ESFAS si as compared to the earlier isolation available via the RPS trip signal (gnal1900 ps Certain other water lines passing through the containment wall are isolated based on different criteria than described in the staff Position Statement at the front of this section. These other lines are described in the following paragraphs.
Reactor Coolant Pump (RCP) Seal Injection Lines As described in NUREG-0623, we have found that an immediate RCP trip must be required at this time for a small break LOCA, and conversely, the RCP's should be kept running for non-LOCA transients. Further we have determined,
that the RCP controls to assure this function should be automatic and of safety grade quality. Although the RCP seal injection system is not an essential system for the inanediate safe (hot) shutdown of the plant, isolation of this system could cause RCP seal damage and a possible LOCA inside containment.
Therefore, we will not pennit automatic isolation of the RCP seal injection water. The seal water supply line can only be isolated by remote mar.al operator action. The valves in this line, requiring operator actica to close, will thus not be subject to spurious automatic valve closure. The seal water return line is automatically isolated on a containment high-high (30 psig) pressure signal.
RCP Motor Cooling Water Our criteria state that all non-essential systems shall be automatically isolated.
We define " essential" systems to include those that nonnally support the operation of the RCP's, and therefore such systems are excluded from the requirement for diverse, automatic isolation. But the RCP motor cooling water may be interrupted
. (Enclosure 3) for 11 definite periods as Icug as seal injectica water (a safe'., 9 aJe system) is being supplied. We therefore approve the RCP motor cooling watc* supply design to continue normal flow after the first containment isolation signals are recieved (4 psig contaicment or 1900 psig reactor coolant). but to isolate after a containment high-high pressure signal (30 psig) is received.
Letdown Line to Purification Deminerslizers The components primarily intended to isolate the letdown line consist of two globe valves in parallel inside containment and a single gate valve outside containment. The globe valves will close on either a containment high pressure signal (4 psig) or an RPS trip signal, which provides at least two diverse closure signals for these valves. The outside gate valve is designed to close automatically only on a containment high pressure signal.
In addition, operator emergency procedures include specific instructions on isolating the letdown line by verifying that certain valves (which are operable from the control room) are closed.
In addition to the tvc globe valves immediately inside containment which are automatically isolated, there are two other parallel globe valves further upstream in the letdown line. These valves are also remotely operable from the control room, providing additional assurance that the letdown line can be isola.ted when necessary.
Having the isolation capability described above, the licensee in ?mendment 12 to the Restart Report has documented a basis for not installing civerse actuating signals to automatically isblate the gate valve in the letdown line immediately outside containment. Their concern is that spurious closure of tiis gate valve would lead to an increase in reactor coolant (due to RCP seal injection flow) of about one to two inches per minute in pressurizer level, which would eventually require termination of RCP seal water injection. This would result in the loss of reactor coolant system inventory control and the primary means of cooling the RCP shaft. The TMI-2 experience showed that post-accident operation of the reacter coolant pumps may be desirable.
In addition, the licensee has stated that letdown line operation is necessary for the controlled plant shutdown to cold shutdown.
Based, on our review, we conclude that post accident operation of the letdown system is important to assurance of a safe, orderly plant shutdown. Accordingly, we will not require diverse isolating signals to the gate valve outside contain-ment, which should reduce the frequency of automatic closure of this vair. and the subsequent probability of the valve failing to open on command when the letdown system function is needed.
d.
O
ENCLOSURE 4
GENERAL COMMENT
S i.
9 Item 2.1.7.a l
So called "CLARf FICATION" in Supplement is in several instances (1, 2, 3, 5, 4, 6a), not clarification al all, is simple restatement in almost identical words - should be eliminated in complete rewrite page C 8-34 Can't tell from Paragraph I what the significance of the parallel /
series arrangement is unless we know whether " contacts" open or close on a failure of RC pump power or MFW pumpgp.
Summary Comnent POSITION should be straight forward statement, itemizing if necessary, but without the subheading called CLARIFICATION.
DISCUSSION should are acceptable for short term, then discuss long term changes or additions to short term changes, explaining how that meets our criteria, too, and is acceptable.
Item 2.1.7.b P 8-7 Suppl.
Said that justification for qual, temps would be required, P C 8-36 St. Rpt.
and the Suppl. names temps of 2500F, 3500F, but does not justify. We should state explicitly why those temps are acceptable. Writeup is too choppy, unfocused on criteria.
Suggest following an outline like this:
Staff Position p
- Safety grade AFW flow indication from sensor to control room
- Diverse electrical power sources to flow-instrument channels in accordance with SRP 10.4.9.
Safety Grade AFW Flow Indication
- Brief system descriptions, level and flow indication
- How the systems, components, meet the following criteria:
Redundancy, single failure, diversity, testability (include tech spec reqmets), seismic & environmental qual.
Diversity of orime power sources.
.